Sample records for uasb reactor system

  1. ANAEROBIC SEWAGE TREATMENT IN A ONE-STAGE UASB AND A COMBINED UASB-DIGESTER SYSTEM

    E-print Network

    Nidal Mahmoud; Grietje Zeeman; Huub Gijzen; Gatze Lettinga

    The potential of a novel technology consisting of a UASB complemented with a digester (UASB-Digester) for mutual sewage treatment and sludge stabilisation under low temperature conditions was investigated. The performance of the UASB-Digester system was compared with a one stage UASB. The UASB reactor was operated at a HRT of 6 hours and controlled temperature of 15°C, the average sewage temperature in the Middle East countries during wintertime, while the digester was operated at 35 °C. The UASB-Digester provided substantially better removal efficiencies and conversion than the one stage UASB reactor (pDigester and the one stage UASB for

  2. Post-treatment of UASB reactor effluent in an integrated duckweed and stabilization pond system

    Microsoft Academic Search

    Peter van der Steen; Asher Brenner; Joost van Buuren; Gideon Oron

    1999-01-01

    Post-treatment of effluent from an Upflow Anaerobic Sludge Blanket (UASB) reactor, that was fed with domestic sewage, was conducted in an integrated pond system. The system consisted of a series of shallow duckweed and stabilization ponds. The main objective of post-treatment is removal of bacterial pathogens and further polishing of effluent quality. Rapid and efficient pathogen removal can be achieved

  3. Slaughterhouse wastewater treatment: evaluation of a new three-phase separation system in a UASB reactor.

    PubMed

    Caixeta, Cláudia E T; Cammarota, Magali C; Xavier, Alcina M F

    2002-01-01

    The anaerobic treatment of the wastewater from the meat processing industry was studied using a 7.2 1 UASB reactor. The reactor was equipped with an unconventional configuration of the three-phase separation system. The effluent was characterized in terms of pH (6.3-6.6), chemical oxygen demand (COD) (2,000-6,000 mg l(-1)), biochemical oxygen demand BOD5 (1,300-2,300 mg 1(-1)), fats (40-600 mg l(-1)) and total suspended solids (TSS) (850-6,300 mg l(-1)) The reactor operated continuously throughout 80 days with hydraulic retention time of 14, 18 and 22 h. The wastewater from Rezende Industrial was collected after it had gone through pretreatment (screening, flotation and equalization). COD, BOD and TSS reductions and the biogas production rate were the parameters considered in analyzing the efficiency of the process. The average production of biogas was 111 day(-1) (STP) for the three experimental runs. COD removal varied from 77% to 91% while BOD removal was 95%. The removal of total suspended solids varied from 81% to 86%. This fact supports optimal efficiency of the proposed three-phase separation system as well as the possibility of applying it to the treatment of industrial effluents. PMID:11710346

  4. UASB reactor hydrodynamics: residence time distribution and proposed modelling tools.

    PubMed

    López, I; Borzacconi, L

    2010-05-01

    The hydrodynamic behaviour of UASB (Up Flow Anaerobic Sludge Blanket) reactors based on residence time distribution curves allows the implementation of global models, including the kinetic aspects of biological reactions. The most relevant hydrodynamic models proposed in the literature are discussed and compared with the extended tanks in series (ETIS) model. Although derived from the tanks in series model, the ETIS model's parameter is not an integer. The ETIS model can be easily solved in the Laplace domain and applied to a two-stage anaerobic digestion linear model. Experimental data from a 250 m3 UASB reactor treating malting wastewater are used to calibrate and validate the proposed model. PMID:20540420

  5. Anaerobic biodegradation of aircraft deicing fluid in UASB reactors.

    PubMed

    Tham, P T Pham thi; Kennedy, K J Kevin J

    2004-05-01

    A central composite design was employed to methodically investigate anaerobic treatment of aircraft deicing fluid (ADF) in bench-scale Upflow Anaerobic Sludge Blanket (UASB) reactors. A total of 23 runs at 17 different operating conditions (0.8% 1.6% ADF (6000-12,000mg/L COD), 12-56h HRT, and 18-36gVSS/L) were conducted in continuous mode. The development of four empirical models describing process responses (i.e. COD removal efficiency, biomass-specific acetoclastic activity, methane production rate, and methane production potential) as functions of ADF concentration, hydraulic retention time, and biomass concentration is presented. Model verification indicated that predicted responses (COD removal efficiencies, biomass-specific acetoclastic activity, and methane production rates and potential) were in good agreement with experimental results. Biomass-specific acetoclastic activity was improved two-fold from 0.23gCOD/gVSS/d for inoculum to a maximum of 0.55gCOD/gVSS/d during ADF treatment in UASB reactors. For the design window, COD removal efficiencies were higher than 90%. The predicted methane production potentials were close to theoretical values, and methane production rates increased as the organic loading rate is increased. ADF toxicity effects were evident for 1.6% ADF at medium organic loadings (SOLR above 0.5gCOD/gVSS/d). In contrast, good reactor stability and excellent COD removal efficiencies were achieved at 1.2% ADF for reactor loadings approaching that of highly loaded systems (0.73gCOD/gVSS/d). PMID:15159155

  6. An examination of the granulation process in UASB reactors

    Microsoft Academic Search

    K. R. K. Alibhai; C. F. Forster

    1986-01-01

    The production of compact, fully stabilized sludge granules is an essential aspect of the operation of UASB reactors. This paper examines how the granulation process is affected by innoculating the seed sludge with solids which had already granulated and by supplementing the feea with calcium and with phosphorous. Of these, phosphorus supplements appear to be the most significant.

  7. UASB reactor hydrodynamics: residence time distribution and proposed modelling tools

    Microsoft Academic Search

    I. López; L. Borzacconi

    2010-01-01

    The hydrodynamic behaviour of UASB (Up Flow Anaerobic Sludge Blanket) reactors based on residence time distribution curves allows the implementation of global models, including the kinetic aspects of biological reactions. The most relevant hydrodynamic models proposed in the literature are discussed and compared with the extended tanks in series (ETIS) model. Although derived from the tanks in series model, the

  8. Maximum COD Loading Capacity in UASB Reactors at 37°C

    Microsoft Academic Search

    Herbert H. P. Fang; H. K. Chui

    1993-01-01

    The maximum capacity chemical oxygen demand (COD) loading in upflow anaerobic sludge blanket (UASB) reactors is evaluated using three 8.5 L reactors and high-strength synthetic wastewaters composed of milk and sucrose at 37~ The study was conducted over a wide-range COD loading rate (18-260 g.L-l.day 1), by varying hydraulic retention time (HRT) (1.8-10 hr) and COD levels in wastewater (6,000-20,000

  9. Performance and behaviour of planted and unplanted units of a horizontal subsurface flow constructed wetland system treating municipal effluent from a UASB reactor.

    PubMed

    da Costa, Jocilene Ferreira; de Paoli, André Cordeiro; Seidl, Martin; von Sperling, Marcos

    2013-01-01

    A system composed of two horizontal subsurface flow constructed wetlands operating in parallel was evaluated for the post-treatment of UASB (upflow anaerobic sludge blanket) reactor effluent, for a population equivalent of 50 inhabitants per unit. One unit was planted with cattail (Typha latifolia) and the other was unplanted. The study was undertaken over a period of 4 years, comprising monitoring of influent and effluent constituents together with a full characterization of the behaviour of the units (tracer studies, mathematical modelling of chemical oxygen demand (COD) decay, characterization of solids in the filter medium). The mean value of the surface hydraulic load was 0.11 m(3)m(-2)d(-1), and the theoretical hydraulic retention time was 1.1 d in each unit. Using tracer tests with (82)Br, dispersion number (d) values of 0.084 and 0.079 for the planted and unplanted units were obtained, indicating low to moderate dispersion. The final effluent had excellent quality in terms of organic matter and suspended solids, but the system showed low capacity for nitrogen removal. Four-year mean effluent concentration values from the planted and unplanted units were, respectively: biochemical oxygen demand (BOD(5)): 25 and 23 mg L(-1); COD: 50 and 55 mg L(-1); total suspended solids (TSS): 9 and 9 mg L(-1); N-ammonia: 27 and 28 mg L(-1). The COD decay coefficient K for the traditional plug-flow model was 0.81 and 0.84 d(-1) for the planted and unplanted units. Around 80% of the total solids present in the filter medium were inorganic, and most of them were present in the interstices rather than attached to the support medium. As an overall conclusion, horizontal subsurface flow wetlands can be a very suitable post-treatment method for municipal effluents from anaerobic reactors. PMID:24135097

  10. Kinetics of concentration decay of specific organic matter in UASB reactors operating with and without return of aerobic sludge.

    PubMed

    Pontes, P P; Chernicharo, C A L; Von Sperling, M

    2014-08-01

    This study aimed at assessing the influence of the return of excess aerobic sludge from a trickling filter (TF) upon the anaerobic digestion process in an upflow anaerobic sludge blanket (UASB) reactor, by evaluating its effect on the kinetics of the decay of specific organic matter (carbohydrates, proteins and lipids), as well as on the concentrations of volatile fatty acids in the UASB reactor. A pilot-scale UASB/TF system was used to perform the experiments, operating with (phase 2) and without (phase 1) excess sludge return from the TF to the UASB reactor. Sampling was carried out at different heights of the UASB reactor (0, 25, 125 and 225-cm height), and profile concentrations were determined for the following parameters: carbohydrates, proteins, lipids and volatile fatty acids. First-order kinetics showed the best fit to the decay of concentrations of carbohydrates, proteins, lipids and chemical oxygen demand (COD) in the UASB reactor. The parameters showing the best fit to the first-order kinetics were proteins and COD, during the sludge return phase. The occurrence of higher apparent reaction constants was further observed during the sludge return phase. For an influent COD concentration of 600 mg L-1 and hydraulic retention times of 2.1, 2.6 and 3.0 h in phase 1, the effluent COD concentrations were 125.3, 88.4 and 62.4 mg L-1, respectively, whereas in phase 2, the effluent COD concentrations were 75.5, 47.6 and 30.1 mg L-1, respectively. PMID:24956799

  11. Microbial community structure of a simultaneous nitrogen and phosphorus removal reactor following treatment in a UASB-DHS system.

    PubMed

    Hatamoto, Masashi; Saito, Yayoi; Dehama, Kazuya; Nakahara, Nozomi; Kuroda, Kyohei; Takahashi, Masanobu; Yamaguchi, Takashi

    2015-01-01

    The anaerobic-anoxic sequence batch reactor (A2SBR) was applied to achieve nitrogen and phosphorus removal in an energy-saving sewage treatment system involving an up-flow anaerobic sludge blanket combined with a down-flow hanging sponge reactor to treat municipal sewage. After sludge acclimation, the A2SBR showed satisfactory denitrification and phosphorus removal performance with total phosphate and nitrate concentrations of the effluent of 8.4 ± 3.4 mg-N L(-1) and 0.9 ± 0.6 mg-P L(-1), respectively. 16S rRNA gene sequence and fluorescence in situ hybridization analyses revealed that 'Candidatus Accumulibacter phosphatis' was the dominant phosphate-accumulating micro-organism. Although a competitive bacterium for polyphosphate-accumulating organisms, 'Ca. Competibacter phosphatis', was not detected, Dechloromonas spp. were abundant. The ppk1 gene sequence analysis showed that the type II lineage of 'Ca. Accumulibacter' was dominant. The results suggest that denitrification and phosphorus removal in the A2SBR could be achieved by cooperative activity of 'Ca. Accumulibacter' and nitrate-reducing bacteria. PMID:25714647

  12. Separation of solids and disinfection for agronomical use of the effluent from a UASB reactor.

    PubMed

    Sundefeld Junior, G C; Piveli, R P; Cutolo, S A; Ferreira Filho, S S; Santos, J G

    2014-01-01

    The present work addresses the preparation of the effluent from a full-scale upflow anaerobic sludge blanket (UASB) reactor for drip irrigation of orange crops. The pilot plant included a lamella plate clarifier followed by a geo-textile blanket filter and a UV disinfection reactor. The clarifier operated with a surface load of 115 m(3)m(-2)d(-1), whereas the filter operated with 10 m(3)m(-2)d(-1). The UV reactor was an open-channel type and the effective dose was approximately 2.8 W h m(-3). The effluent of the UASB reactor received 0.5 mg L(-1) cationic polyelectrolyte before entering the high-rate clarifier. Suspended solids' concentrations and Escherichia coli and helminth egg's densities were monitored throughout the treatment system for 12 months. Results showed that the total suspended solids concentration in the filter effluent was lower than 7 mg L(-1) and helminth density was below 1.0 egg L(-1). The UV disinfection demonstrated the ability to produce a final effluent with E. coli density lower than 10(3)MPN/100 mL (MPN: most probable number) during the entire process. Thus, the World Health Organization standards for unrestricted crop use were met. Agronomic interest parameters were controlled and it was possible to identify the important contribution of treated sewage in terms of the main nutrients. PMID:24434964

  13. Enhanced reductive transformation of p-chloronitrobenzene in a novel bioelectrode-UASB coupled system.

    PubMed

    Zhu, Liang; Gao, Kaituo; Qi, Jiaoqin; Jin, Jie; Xu, Xiangyang

    2014-09-01

    The laboratory-scale upflow anaerobic sludge blanket (UASB) reactor equipped with a pair of bioelectrodes was established for the enhancement of p-chloronitrobenzene (p-ClNB) reductive transformation via the electrolysis. Results showed that a stable COD removal efficiency over 99% and high p-ClNB transformation rate of 0.328 h(-1) were achieved in the bioelectrode-UASB coupled system with influent COD and p-ClNB loading rates of 2.1-4.2 kg COD m(-3)d(-1) and 60 gm(-3)d(-1), respectively. The bioelectrodes were supplied with a voltage of 2.5-5.0 V and the effective current was above 2 mA, which resulted in a continuous supply of H2. Compared with the traditional UASB reactor (R1), the production of H2 was promoted in the bioelectrode-UASB coupled system (R2), and was consumed as an internal electron donor for p-ClNB reductive transformation by anaerobic microbes simultaneously. Furthermore, the cyclic voltammetry curve (CV) analysis of biocathodes showed a positive shift in the reductive peak potential and a dramatic increase in the reductive peak current, which demonstrated the catalytic reduction of p-ClNB by biocathode in the combined system. PMID:24997372

  14. Use of UASB reactors for brackish aquaculture sludge digestion under different conditions.

    PubMed

    Mirzoyan, Natella; Gross, Amit

    2013-05-15

    Treatment and disposal of high volume of salty waste production in recirculating aquaculture systems (RASs) is a major challenge and the sludge is often a source of environmental pollution and salinization of receiving soils and water bodies. Anaerobic digestion is an efficient mean for the treatment of wastes of different origins and might serve a useful tool for the reduction of salty aquaculture discharge load. Use of an upflow anaerobic sludge blanket (UASB) reactor for digestion of brackish aquaculture sludge from RASs under different C:N ratios, temperatures, and hydraulic retention times demonstrated high removal efficiencies of over 92% as volatile solids (VS), 98% as chemical oxygen demand and 81% as total suspended solids in all reactors. Methane production topped 7.1 mL/gVS d and was limited by low C:N ratio but was not influenced by temperature fluctuations. The treated liquid effluent from all reactors was of sufficient quality for reuse in the RAS, leading to significant water recycling and saving rates. UASB may be an attractive solution for brackish sludge management in RASs. PMID:23528783

  15. Nutrient recovery from domestic wastewater using a UASB-duckweed ponds system

    Microsoft Academic Search

    Saber A. El-Shafai; Fatma A. El-Gohary; Fayza A. Nasr; N. Peter van der Steen; Huub J. Gijzen

    2007-01-01

    The pilot-scale wastewater treatment system used in this study comprised a 40-l UASB reactor (6-h HRT) followed by three duckweed ponds in series (total HRT 15days). During the warm season, the treatment system achieved removal values of 93%, 96% and 91% for COD, BOD and TSS, respectively. Residual values of ammonia, TKN and total phosphorus were 0.41mg N\\/l, 4.4mg N\\/l

  16. The UASB reactor as an alternative for the septic tank for on-site sewage treatment.

    PubMed

    Coelho, A L S S; do Nascimento, M B H; Cavalcanti, P F F; van Haandel, A C

    2003-01-01

    Although septic tanks are amply used for on site sewage treatment, these units have serious drawbacks: the removal efficiency of organic material and suspended solids is low, the units are costly and occupy a large area and operational cost is high due to the need for periodic desludging. In this paper an innovative variant of the UASB reactor is proposed as an alternative for the septic tank. This alternative has several important advantages in comparison with the conventional septic tank: (1) Although the volume of the UASB reactor was about 4 times smaller than the septic tank, its effluent quality was superior, even though small sludge particles were present, (2) desludging of the UASB reactor is unnecessary and even counterproductive, as the sludge mass guarantees proper performance, (3) the UASB reactor is easily transportable (compact and light) and therefore can be produced in series, strongly reducing construction costs and (4) since the concentration of colloids in the UASB effluent is much smaller than in the ST effluent, it is expected that the infiltration of the effluent will be much less problematic. PMID:14753540

  17. Anaerobic treatment of brewery wastewater using UASB reactors seeded with activated sludge

    Microsoft Academic Search

    C. Cronin; K. V. Lo

    1998-01-01

    In this study, two upflow anaerobic sludge blanket (UASB) reactors seeded with aerobic activated sludge were used to treat brewery wastewater. The reactors were identical in design and were continuously operated at 19–24°C. Reactor A was seeded with 1·98 g volatile suspended solids (VSS) per liter of acclimatized activated sludge, while Reactor B was seeded with 1·98 g VSS l?1

  18. Treatment of slaughterhouse wastewater in a UASB reactor and an anaerobic filter

    Microsoft Academic Search

    I. Ruiz; M. C. Veiga; P. de Santiago; R. Blázquez

    1997-01-01

    A study was performed to assess the feasibility of anaerobic treatment of slaughterhouse wastewaters in a UASB (Upflow Anaerobic Sludge Blanket) reactor and in an AF (Anaerobic Filter). Among the different streams generated, the slaughter line showed the highest organic content with an average COD of 8000 mg\\/l, of which 70% was proteins. The suspended solids content represented between 15

  19. Comparison of UASB and EGSB reactors performance, for treatment of raw and deoiled palm oil mill effluent (POME).

    PubMed

    Fang, Cheng; O-Thong, Sompong; Boe, Kanokwan; Angelidaki, Irini

    2011-05-15

    Anaerobic digestion of palm oil mill effluent (POME) and deoiled POME was investigated both in batch assays and continuous reactor experiments using up-flow anaerobic sludge blanket (UASB) and expanded granular sludge bed (EGSB) reactors. The methane potential determined from batch assays of POME and deoiled POME was 503 and 610 mL-CH(4)/gVS-added, respectively. For the treatment of POME in continuously fed reactors, both in UASB and EGSB reactors more than 90% COD removal could be obtained, at HRT of 5 days, corresponding to OLR of 5.8 gVS/(L-reactor.d). Similar methane yields of 436-438 mL-CH(4)/gVS-added were obtained for UASB and EGSB respectively. However, for treatment of deoiled POME, both UASB and EGSB reactors could operate at lower OLR of 2.6 gVS/(L-reactor.d), with the methane yield of 600 and 555 mL-CH(4)/gVS-added for UASB and EGSB, respectively. The higher methane yield achieved from the deoiled POME was attributed to lower portion of biofibers which are more recalcitrant compared the rest of organic matter in POME. The UASB reactor was found to be more stable than EGSB reactor under the same OLR, as could be seen from lower VFA concentration, especially propionic acid, compared to the EGSB reactor. PMID:21377272

  20. Nitrification-denitrification of UASB effluents highly loaded with nitrogen in an activated sludge reactor operated with short cycled aeration

    Microsoft Academic Search

    S. Villaverde; M. L. Lacalle; P. A. García-Encina; F. Fdz-Polanco

    2001-01-01

    A conventional activated sludge reactor operated with short cycled aeration was used for total nitrogen removal of UASB anaerobic reactor effluent containing nitrogen (up to 1,200 mg NKT\\/L) and organic matter (up to 2,000 mg COD\\/L). Initially the reactor was fed with synthetic water to progressively introduce the UASB effluent. This favored the acclimation of the microorganisms to the real

  1. Comparative performance of a UASB reactor and an anaerobic packed-bed reactor when treating potato waste leachate

    Microsoft Academic Search

    W. Parawira; M. Murto; R. Zvauya; B. Mattiasson

    2006-01-01

    The results presented in this paper are from studies on a laboratory-scale upflow anaerobic sludge blanket (UASB) reactor and an anaerobic packed-bed (APB) reactor treating potato leachate at increasing organic loading rates from 1.5 to 7.0gCOD\\/1\\/day. The hydraulic retention times ranged from 13.2 to 2.8 days for both reactors during the 100 days of the experiment. The maximum organic loading

  2. Performance and kinetics of an upflow anaerobic sludge blanket (UASB) reactor treating slaughterhouse wastewater

    Microsoft Academic Search

    R. Borja; C. J. Banks; Z. Wang

    1994-01-01

    A laboratory study was devised to verify the use of Monod kinetics for the modelling of a continuous upflow anaerobic sludge blanket (UASB) reactor treating slaughterhouse wastewater. Eight different hydraulic retention times (4.0–0.3 day) were investigated at an average influent COD concentration of 5.1 g.l. The maximum substrate utilization rate, k, and half saturation coefficient, KL, were determined to be

  3. Effect of sulfate on anaerobic degradation of benzoate in UASB reactors

    SciTech Connect

    Fang, H.H.P.; Liu, Y.; Chen, T. [Univ. of Hong Kong (Hong Kong)

    1997-04-01

    Anaerobic processes have been widely used for the treatment of various high-strength industrial wastewaters. However, application has been limited for the treatment of sulfate-rich industrial wastewaters, such as those from the petrochemical, and mining industries. Wastewaters containing benzoate and sulfate were treated in two upflow anaerobic sludge blanket (UASB) reactors at 34--37 C for 320 d. The sulfate concentration was increased stepwise in Reactor-A up to 7,500 mg/L, and was kept mostly constant at 3,000 mg/L in Reactor-B. Both reactors removed over 98% of organic chemical-oxygen demand (COD) for sulfate up to 6,000 mg/L, despite the fact that the mixed liquor contained up to 769 mg S/L of total sulfides and up to 234 mg S/L of dissolved H{sub 2}S. Sulfate0reducing efficiency decreased with the increase in sulfate concentration, but increased with time at each sulfate concentration. Reactor-B consistently reduced 89% of sulfate. However, both organic COD removal and sulfate-reducing efficiencies of Reactor-A dropped drastically at 7,500 mg SO{sub 4}{sup {minus}2}/L, and showed no sign of recovery after 50 d. The system failure was likely due to the increased sulfate, instead of sulfide, toxicity. From the COD balance, 93.4% of COD removed was converted to methane instead of sulfides, with a net sludge yield of 0.047 g volatile suspended solids (VSS)/g COD. The sulfur balance was over 97%.

  4. Shift of propionate-oxidizing bacteria with HRT decrease in an UASB reactor containing propionate as a sole carbon source.

    PubMed

    Ban, Qiaoying; Zhang, Liguo; Li, Jianzheng

    2015-01-01

    Propionate is a main intermediate product, and its degradation is crucial for maintaining the efficiency and stability of an anaerobic reactor. However, there was little information about the effects of ecological factor on propionate-oxidizing bacteria. In current research, microbial community composition and quantitative analysis of some identified propionate-oxidizing bacteria with hydraulic retention time (HRT) decrease in an upflow anaerobic sludge blanket (UASB) reactor containing propionate as sole carbon source was investigated. The results showed that propionate-oxidizing bacteria from Syntrophobacter, Pelotomaculum, and Smithella were major functional bacteria in this UASB system. Most propionate-oxidizing bacteria in composition have not changed with HRT decrease. However, the number of previously identified propionate-oxidizing bacteria from these three genera exhibited significant shift. Under HRT 10 h condition, Pelotomaculum schinkii was dominant and its quantity was 1.2?×?10(4) 16S ribosomal RNA (rRNA) gene copies/ng DNA, occupying 56.2 % in total detectable propionate-oxidizing bacteria. HRT decrease from 10 h to 8 and 6 h stepwise resulted in P. schinkii, Syntrophobacter sulfatireducens and Smithella propionica becoming the main population. HRT decrease from 6 to 4 h did not markedly change the amount of propionate-oxidizing bacteria, but S. propionica dominated in the reactor. PMID:25261998

  5. Performance evaluation of planted and unplanted subsurface-flow constructed wetlands for the post-treatment of UASB reactor effluents.

    PubMed

    Dornelas, Filipe Lima; Machado, Matheus Boechat; von Sperling, Marcos

    2009-01-01

    A system comprised by a UASB (Upflow Anaerobic Sludge Blanket) reactor followed by two horizontal subsurface-flow constructed wetlands in parallel was evaluated for the treatment of the wastewater generated in the city of Belo Horizonte, Brazil (50 inhabitants each unit). One unit was planted (Typha latifolia) and the other was unplanted. Influent and effluent samples were collected for a period of seven months. The systems were able to produce final effluents with low concentrations of organic matter and suspended solids, but showed not to be efficient in the removal of nutrients. Mean effluent concentrations for the planted and unplanted units were, respectively: BOD: 15 and 19 mg/L; COD: 42 and 64 mg/L; TSS: 3 and 5 mg/L; TN: 27 and 33 mg/L; N-NH(3): 25 and 29 mg/L; P Total: 1.2 and 1.5 mg/L. The planted wetland presented effluent concentrations and removal efficiencies significantly (Wilcoxon matched-pairs test, 5% significance level) better than the unplanted unit for most constituents. The study shows that horizontal subsurface-flow constructed wetlands can be effectively used as a post-treatment option for the effluent from UASB reactors. PMID:19955625

  6. Understanding the removal mechanisms of PPCPs and the influence of main technological parameters in anaerobic UASB and aerobic CAS reactors.

    PubMed

    Alvarino, T; Suarez, S; Lema, J M; Omil, F

    2014-08-15

    The removal of 16 Pharmaceutical and Personal Care Products (PPCPs) were studied in a conventional activated sludge (CAS) unit and an upflow anaerobic sludge blanket (UASB) reactor. Special attention was paid to each biomass conformation and activity as well as to operational conditions. Biodegradation was the main PPCP removal mechanism, being higher removals achieved under aerobic conditions, except in the case of sulfamethoxazole and trimetrophim. Under anaerobic conditions, PPCP biodegradation was correlated with the methanogenic rate, while in the aerobic reactor a relationship with nitrification was found. Sorption onto sludge was influenced by biomass conformation, being only significant for musk fragrances in the UASB reactor, in which an increase of the upward velocity and hydraulic retention time improved this removal. Additionally, PPCP sorption increased with time in the UASB reactor, due to the granular biomass structure which suggests the existence of intra-molecular diffusion. PMID:25010455

  7. Performance assessment of different STPs based on UASB followed by aerobic post treatment systems

    PubMed Central

    2014-01-01

    This paper present the experiences gained from the study of ten up flow anaerobic sludge blanket (UASB) based sewage treatment plants (STPs) of different cities of India. Presently 37 UASB based STPs were under operation and about 06 UASB based STPs are under construction and commissioning phase at different towns. The nature of sewage significantly varied at each STP. Two STP were receiving sewage with high sulfate and heavy metals due to the mixing of industrial waste. The treatment performance of all UASB reactors in terms of BOD, COD and TSS were observed between 55 to 70% respectively. The post treatment units down flow hanging sponge (DHS) and Aeration followed by activated sludge process (ASP) at two STPs were performing well and enable to achieve the required disposal standards. Results indicate the effluent quality in terms of BOD and SS were less than 30 and 50 mg/L and well below the discharging standards. PMID:24468307

  8. Microbial dynamics during azo dye degradation in a UASB reactor supplied with yeast extract.

    PubMed

    Silva, S Q; Silva, D C; Lanna, M C S; Baeta, B E L; Aquino, S F

    2014-01-01

    The present work aimed to investigate the microbial dynamics during the anaerobic treatment of the azo dye blue HRFL in bench scale upflow anaerobic sludge bed (UASB) reactor operated at ambient temperature. Sludge samples were collected under distinct operational phases, when the reactor were stable (low variation of color removal), to assess the effect of glucose and yeast extract as source of carbon and redox mediators, respectively. Reactors performance was evaluated based on COD (chemical oxygen demand) and color removal. The microbial dynamics were investigated by PCR-DGGE (Polimerase Chain Reaction - Denaturing Gradient of Gel Electrophoresis) technique by comparing the 16S rDNA profiles among samples. The results suggest that the composition of microorganisms changed from the beginning to the end of the reactor operation, probably in response to the presence of azo dye and/or its degradation byproducts. Despite the highest efficiency of color removal was observed in the presence of 500 mg/L of yeast extract (up to 93%), there were no differences regarding the microbial profiles that could indicate a microbial selection by the yeast extract addition. On the other hand Methosarcina barkeri was detected only in the end of operation when the best efficiencies on color removal occurred. Nevertheless the biomass selection observed in the last stages of UASB operation is probably a result of the washout of the sludge in response of accumulation of aromatic amines which led to tolerant and very active biomass that contributed to high efficiencies on color removal. PMID:25763018

  9. Microbial dynamics during azo dye degradation in a UASB reactor supplied with yeast extract

    PubMed Central

    Silva, S.Q.; Silva, D.C.; Lanna, M.C.S.; Baeta, B.E.L.; Aquino, S.F.

    2014-01-01

    The present work aimed to investigate the microbial dynamics during the anaerobic treatment of the azo dye blue HRFL in bench scale upflow anaerobic sludge bed (UASB) reactor operated at ambient temperature. Sludge samples were collected under distinct operational phases, when the reactor were stable (low variation of color removal), to assess the effect of glucose and yeast extract as source of carbon and redox mediators, respectively. Reactors performance was evaluated based on COD (chemical oxygen demand) and color removal. The microbial dynamics were investigated by PCR-DGGE (Polimerase Chain Reaction - Denaturing Gradient of Gel Electrophoresis) technique by comparing the 16S rDNA profiles among samples. The results suggest that the composition of microorganisms changed from the beginning to the end of the reactor operation, probably in response to the presence of azo dye and/or its degradation byproducts. Despite the highest efficiency of color removal was observed in the presence of 500 mg/L of yeast extract (up to 93%), there were no differences regarding the microbial profiles that could indicate a microbial selection by the yeast extract addition. On the other hand Methosarcina barkeri was detected only in the end of operation when the best efficiencies on color removal occurred. Nevertheless the biomass selection observed in the last stages of UASB operation is probably a result of the washout of the sludge in response of accumulation of aromatic amines which led to tolerant and very active biomass that contributed to high efficiencies on color removal. PMID:25763018

  10. Simple wastewater treatment (UASB reactor, shallow polishing ponds, coarse rock filter) allowing compliance with different reuse criteria.

    PubMed

    von Sperling, M; de Andrada, J G B

    2006-01-01

    UASB reactors followed by polishing ponds comprise simple and economic wastewater treatment systems, capable of reaching very high removal efficiencies of pathogenic organisms, leading to the potential use of the effluent for unrestricted irrigation. However, for other types of reuse (urban and industrial), ponds are limited in the sense of producing effluents with high suspended solids (algae) concentrations. The work investigates a system with coarse rock filters for polishing the pond effluent. The overall performance of the system is analyzed, together with the potential for different types of reuse. The excellent results obtained (mean effluent concentrations: BOD: 27 mg/L; SS: 26 mg/L; E. coli: 450 MPN/100 mL) indicate the possibility of unrestricted use of the effluent for agriculture and restricted urban and industrial uses, according to WHO and USEPA. PMID:17302321

  11. Comparison of UASB and EGSB reactors performance, for treatment of raw and deoiled palm oil mill effluent (POME)

    Microsoft Academic Search

    Cheng Fang; Sompong O-Thong; Kanokwan Boe; Irini Angelidaki

    2011-01-01

    Anaerobic digestion of palm oil mill effluent (POME) and deoiled POME was investigated both in batch assays and continuous reactor experiments using up-flow anaerobic sludge blanket (UASB) and expanded granular sludge bed (EGSB) reactors. The methane potential determined from batch assays of POME and deoiled POME was 503 and 610mL-CH4\\/gVS-added, respectively. For the treatment of POME in continuously fed reactors,

  12. Anaerobic degradation of aircraft deicing fluid (ADF) in upflow anaerobic sludge blanket (UASB) reactors and the fate of ADF additives

    NASA Astrophysics Data System (ADS)

    Pham, Thi Tham

    2002-11-01

    A central composite design was employed to methodically investigate anaerobic treatment of aircraft deicing fluid (ADF) in bench-scale Upflow Anaerobic Sludge Blanket (UASB) reactors. A total of 23 runs at 17 different operating conditions were conducted in continuous mode. The development of four empirical models describing process responses (i.e., chemical oxygen demand (COD) removal efficiency, biomass specific acetoclastic activity, methane production rate, and methane production potential) as functions of ADF concentration, hydraulic retention time (HRT), and biomass concentration is presented. Model verification indicated that predicted responses (COD removal efficiencies, biomass specific acetoclastic activity, and methane production rates and potential) were in good agreement with experimental results. Biomass specific acetoclastic activity was improved by almost two-fold during ADF treatment in UASB reactors. For the design window, COD removal efficiencies were higher than 90%. Predicted methane production potentials were close to theoretical values, and methane production rates increased as the organic loading rate (OLR) was increased. ADF toxicity effects were evident for 1.6% ADF at medium specific organic loadings (SOLR above 0.5 g COD/g VSS/d). In contrast, good reactor stability and excellent removal efficiencies were achieved at 1.2% ADF for reactor loadings approaching that of highly loaded systems (0.73 g COD/g VSS/d). Acclimation to ADF resulted in an initial reduction in the biomass settling velocity. The fate of ADF additives was also investigated. There was minimal sorption of benzotriazole (BT), 5-methyl-1 H-benzotriazole (MeBT), and 5,6-dimethyl-1 H-benzotriazole (DiMeBT) to anaerobic granules. A higher sorption capacity was measured for NP. Active transport may be one of the mechanisms for NP sorption. Ethylene glycol degradation experiments indicated that BT, MeBT, DiMeBT, and the nonionic surfactant Tergitol NP-4 had no significant effects on acidogenesis and methanogenesis at the concentration levels studied. A significant inhibition of acetoclastic activity was observed for NP at 100 mg/L, with acetic acid consumption rate at 38% of that for controls. No evidence for anaerobic degradation of benzotriazole and its derivatives was observed; however, both batch and continuous experiments suggested that anaerobic degradation of NP occurred. Kinetic analysis of operational data obtained for the anaerobic treatment of ADF in UASB reactors indicated that the substrate utilization rate was independent of the reactor biomass concentration. The maximum rate of substrate utilization and the half-velocity constants for ADF treatment were 28.4 g COD/L/d and 648 mg COD/L, respectively. For 1.2% ADF, the biomass yield and endogenous decay coefficients were 0.027 g VSS/g COD and 0.012 d-1 , respectively.

  13. Role of calcium oxide in sludge granulation and methanogenesis for the treatment of palm oil mill effluent using UASB reactor.

    PubMed

    Ahmad, Anwar; Ghufran, Rumana; Abd Wahid, Zularisam

    2011-12-30

    The granulation process in palm oil mill effluent using calcium oxide-cement kiln dust (CaO-CKD) provides an attractive and cost effective treatment option. In this study the efficiency of CaO-CKD at doses of 1.5-20 g/l was tested in batch experiments and found that 10 g of CaO/l caused the greatest degradation of VFA, butyrate and acetate. An upflow anaerobic sludge blanket (UASB) reactor was operated continuously at 35°C for 150 days to investigate the effect of CaO-CKD on sludge granulation and methanogenesis during start-up. The treatment of POME emphasized the influence of varying organic loading rates (OLR). Up to 94.9% of COD was removed when the reactor was fed with the 15.5-65.5 g-CODg/l at an OLR of 4.5-12.5 kg-COD/m(3)d, suggesting the feasibility of using CaO in an UASB process to treat POME. The ratio of volatile solids/total solids (VS/TS) and volatile fatty acids in the anaerobic sludge in the UASB reactor decreased significantly after long-term operation due to the precipitation of calcium carbonate in the granules. Granulation and methanogenesis decreased with an increase in the influent CaO-CKD concentration. PMID:22047724

  14. Biological sulfate removal from acrylic fiber manufacturing wastewater using a two-stage UASB reactor.

    PubMed

    Li, Jin; Wang, Jun; Luan, Zhaokun; Ji, Zhongguang; Yu, Lian

    2012-01-01

    A two-stage UASB reactor was employed to remove sulfate from acrylic fiber manufacturing wastewater. Mesophilic operation (35 +/- 0.5 degree C) was performed with hydraulic retention time (HRT) varied between 28 and 40 hr. Mixed liquor suspended solids (MLSS) in the reactor was maintained about 8000 mg/L. The results indicated that sulfate removal was enhanced with increasing the ratio of COD/SO4(2-). At low COD/SO4(2-), the growth of the sulfate-reducing bacteria (SRB) was carbon-limited. The optimal sulfate removal efficiencies were 75% when the HRT was no less than 38 hr. Sulfidogenesis mainly happened in the sulfate-reducing stage, while methanogenesis in the methane-producing stage. Microbes in sulfate-reducing stage performed granulation better than that in methane-producing stage. Higher extracellular polymeric substances (EPS) content in sulfate-reducing stage helped to adhere and connect the flocculent sludge particles together. SRB accounted for about 31% both in sulfate-reducing stage and methane-producing stage at COD/SO4(2-) ratio of 0.5, while it dropped dramatically from 34% in sulfate-reducing stage to 10% in methane-producing stage corresponding to the COD/SO4(2-) ratio of 4.7. SRB and MPA were predominant in sulfate-reducing stage and methane-producing stage respectively. PMID:22655398

  15. A fuzzy-logic-based model to predict biogas and methane production rates in a pilot-scale mesophilic UASB reactor treating molasses wastewater.

    PubMed

    Turkdogan-Aydinol, F Ilter; Yetilmezsoy, Kaan

    2010-10-15

    A MIMO (multiple inputs and multiple outputs) fuzzy-logic-based model was developed to predict biogas and methane production rates in a pilot-scale 90-L mesophilic up-flow anaerobic sludge blanket (UASB) reactor treating molasses wastewater. Five input variables such as volumetric organic loading rate (OLR), volumetric total chemical oxygen demand (TCOD) removal rate (R(V)), influent alkalinity, influent pH and effluent pH were fuzzified by the use of an artificial intelligence-based approach. Trapezoidal membership functions with eight levels were conducted for the fuzzy subsets, and a Mamdani-type fuzzy inference system was used to implement a total of 134 rules in the IF-THEN format. The product (prod) and the centre of gravity (COG, centroid) methods were employed as the inference operator and defuzzification methods, respectively. Fuzzy-logic predicted results were compared with the outputs of two exponential non-linear regression models derived in this study. The UASB reactor showed a remarkable performance on the treatment of molasses wastewater, with an average TCOD removal efficiency of 93 (+/-3)% and an average volumetric TCOD removal rate of 6.87 (+/-3.93) kg TCOD(removed)/m(3)-day, respectively. Findings of this study clearly indicated that, compared to non-linear regression models, the proposed MIMO fuzzy-logic-based model produced smaller deviations and exhibited a superior predictive performance on forecasting of both biogas and methane production rates with satisfactory determination coefficients over 0.98. PMID:20609515

  16. Coagulant recovery from water treatment plant sludge and reuse in post-treatment of UASB reactor effluent treating municipal wastewater.

    PubMed

    Nair, Abhilash T; Ahammed, M Mansoor

    2014-09-01

    In the present study, feasibility of recovering the coagulant from water treatment plant sludge with sulphuric acid and reusing it in post-treatment of upflow anaerobic sludge blanket (UASB) reactor effluent treating municipal wastewater were studied. The optimum conditions for coagulant recovery from water treatment plant sludge were investigated using response surface methodology (RSM). Sludge obtained from plants that use polyaluminium chloride (PACl) and alum coagulant was utilised for the study. Effect of three variables, pH, solid content and mixing time was studied using a Box-Behnken statistical experimental design. RSM model was developed based on the experimental aluminium recovery, and the response plots were developed. Results of the study showed significant effects of all the three variables and their interactions in the recovery process. The optimum aluminium recovery of 73.26 and 62.73 % from PACl sludge and alum sludge, respectively, was obtained at pH of 2.0, solid content of 0.5 % and mixing time of 30 min. The recovered coagulant solution had elevated concentrations of certain metals and chemical oxygen demand (COD) which raised concern about its reuse potential in water treatment. Hence, the coagulant recovered from PACl sludge was reused as coagulant for post-treatment of UASB reactor effluent treating municipal wastewater. The recovered coagulant gave 71 % COD, 80 % turbidity, 89 % phosphate, 77 % suspended solids and 99.5 % total coliform removal at 25 mg Al/L. Fresh PACl also gave similar performance but at higher dose of 40 mg Al/L. The results suggest that coagulant can be recovered from water treatment plant sludge and can be used to treat UASB reactor effluent treating municipal wastewater which can reduce the consumption of fresh coagulant in wastewater treatment. PMID:24777321

  17. Quantification of organic and nitrogen removal in downflow hanging sponge (DHS) systems as a post-treatment of UASB effluent.

    PubMed

    Wichitsathian, B; Racho, P

    2010-01-01

    The aim of this research was to investigate the nature and composition of organic substrate in two down-flow hanging sponge (DHS) systems using mixed fungal (FDHS) and bacterial (BDHS) cultures treatment for UASB effluent of tapioca starch wastewater, evaluated by COD fractionations and two material balances. The random type DHS reactors were operated as modular columns consisting of four identical segments connected vertically. Results of the wastewater characterization showed that carbonaceous fractions were varied on a function of DHS height. Two balances applied to experimental data were for chemical oxygen demand (COD) and nitrogen (N). Results of mass balance calculations can also be used to examine the process behavior of two DHS systems to improve the organic and nitrogen removal mechanisms. PMID:21045340

  18. Biotic and abiotic processes contribute to successful anaerobic degradation of cyanide by UASB reactor biomass treating brewery waste water.

    PubMed

    Novak, Domen; Franke-Whittle, Ingrid H; Pirc, Elizabeta Tratar; Jerman, Vesna; Insam, Heribert; Logar, Romana Marinšek; Stres, Blaž

    2013-07-01

    In contrast to the general aerobic detoxification of industrial effluents containing cyanide, anaerobic cyanide degradation is not well understood, including the microbial communities involved. To address this knowledge gap, this study measured anaerobic cyanide degradation and the rearrangements in bacterial and archaeal microbial communities in an upflow anaerobic sludge blanket (UASB) reactor biomass treating brewery waste water using bio-methane potential assays, molecular profiling, sequencing and microarray approaches. Successful biogas formation and cyanide removal without inhibition were observed at cyanide concentrations up to 5 mg l(-1). At 8.5 mg l(-1) cyanide, there was a 22 day lag phase in microbial activity, but subsequent methane production rates were equivalent to when 5 mg l(-1) was used. The higher cumulative methane production in cyanide-amended samples indicated that part of the biogas was derived from cyanide degradation. Anaerobic degradation of cyanide using autoclaved UASB biomass proceeded at a rate more than two times lower than when UASB biomass was not autoclaved, indicating that anaerobic cyanide degradation was in fact a combination of simultaneous abiotic and biotic processes. Phylogenetic analyses of bacterial and archaeal 16S rRNA genes for the first time identified and linked the bacterial phylum Firmicutes and the archaeal genus Methanosarcina sp. as important microbial groups involved in cyanide degradation. Methanogenic activity of unadapted granulated biomass was detected at higher cyanide concentrations than reported previously for the unadapted suspended biomass, making the aggregated structure and predominantly hydrogenotrophic nature of methanogenic community important features in cyanide degradation. The combination of brewery waste water and cyanide substrate was thus shown to be of high interest for industrial level anaerobic cyanide degradation. PMID:23726700

  19. Modified ADM1 for modelling an UASB reactor laboratory plant treating starch wastewater and synthetic substrate load tests.

    PubMed

    Hinken, L; Huber, M; Weichgrebe, D; Rosenwinkel, K-H

    2014-11-01

    A laboratory plant consisting of two UASB reactors was used for the treatment of industrial wastewater from the wheat starch industry. Several load tests were carried out with starch wastewater and the synthetic substrates glucose, acetate, cellulose, butyrate and propionate to observe the impact of changing loads on gas yield and effluent quality. The measurement data sets were used for calibration and validation of the Anaerobic Digestion Model No. 1 (ADM1). For a precise simulation of the detected glucose degradation during load tests with starch wastewater and glucose, it was necessary to incorporate the complete lactic acid fermentation into the ADM1, which contains the formation and degradation of lactate and a non-competitive inhibition function. The modelling results of both reactors based on the modified ADM1 confirm an accurate calculation of the produced gas and the effluent concentrations. Especially, the modelled lactate effluent concentrations for the load cases are similar to the measurements and justified by literature. PMID:25043796

  20. Diversity and dynamics of ammonia-oxidizing bacterial communities in a sponge-based trickling filter treating effluent from a UASB reactor.

    PubMed

    Mac Conell, E F A; Almeida, P G S; Zerbini, A M; Brandt, E M F; Araújo, J C; Chernicharo, C A L

    2013-01-01

    Changes in ammonia-oxidizing bacterial (AOB) population dynamics were examined in a new sponge-based trickling filter (TF) post-upflow anaerobic sludge blanket (UASB) reactor by denaturating gradient gel electrophoresis (DGGE), and these changes were linked to relevant components influencing nitrification (chemical oxygen demand (COD), nitrogen (N)). The sponge-based packing media caused strong concentration gradients along the TF, providing an ecological selection of AOB within the system. The organic loading rate (OLR) affected the population dynamics, and under higher OLR or low ammonium-nitrogen (NH4(+)-N) concentrations some AOB bands disappeared, but maintaining the overall community function for NH4(+)-N removal. The dominant bands present in the upper portions of the TF were closely related to Nitrosomonas europaea and distantly affiliated to Nitrosomonas eutropha, and thus were adapted to higher NH4(+)-N and organic matter concentrations. In the lower portions of the TF, the dominant bands were related to Nitrosomonas oligotropha, commonly found in environments with low levels of NH4(+)-N. From a technology point of view, changes in AOB structure at OLR around 0.40-0.60 kgCOD m(-3) d(-1) did not affect TF performance for NH4(+)-N removal, but AOB diversity may have been correlated with the noticeable stability of the sponge-based TF for NH4(+)-N removal at low OLR. This study is relevant because molecular biology was used to observe important features of a bioreactor, considering realistic operational conditions applied to UASB/sponge-based TF systems. PMID:23925194

  1. Comparison between polishing (maturation) ponds and subsurface flow constructed wetlands (planted and unplanted) for the post-treatment of the effluent from UASB reactors.

    PubMed

    von Sperling, M; Dornelas, F L; Assunção, F A L; de Paoli, A C; Mabub, M O A

    2010-01-01

    This paper presents the results of a comparison of the performance of two treatment systems operating in parallel, with the same influent wastewater. The investigated systems are (i) UASB + three polishing ponds in series + coarse filter (200 population equivalents) and (ii) UASB + subsurface flow constructed wetlands (50 population equivalents). Two wetland units, operating in parallel, were analysed, being one planted (Typha latifolia) and the other unplanted. The systems were located in Belo Horizonte, Brazil. The wetland systems showed to be more efficient in the removal of organic matter and suspended solids, leading to good effluent BOD and COD concentrations and excellent SS concentrations. The planted wetland performed better than the unplanted unit, but the latter was also able to provide a good effluent quality. The polishing pond system was more efficient in the removal of nitrogen (ammonia) and coliforms (E. coli). Land requirements and cost considerations are presented. PMID:20220242

  2. Microbial community composition and reactor performance during hydrogen production in a UASB reactor fed with raw cheese whey inoculated with compost.

    PubMed

    Castelló, E; Perna, V; Wenzel, J; Borzacconi, L; Etchebehere, C

    2011-01-01

    This study investigated the microbial community developed in a UASB reactor for hydrogen production and correlated it to reactor performance. The reactor was inoculated with kitchen waste compost and fed with raw cheese whey at two organic loading rates, 20 gCOD/Ld and 30 gCOD/Ld. Hydrogen production was very variable, using an OLR of 30 gCOD/Ld averaged 1.0 LH(2)/Ld with no methane produced under these conditions. The hydrogen yield was also very variable and far from the theoretical. This low yield could be explained by selection of a mixed fermentative population with presence of hydrogen producing organisms (Clostridium, Ruminococcus and Enterobacter) and other non-hydrogen producing fermenters (Lactobacillus, Dialister and Prevotella). The molecular analysis of the raw cheese whey used for feeding revealed the presence of three predominant organisms that are affiliated with the genera Buttiauxella (a low-yield hydrogen producer) and Streptococcus (a lactic acid-producing fermenter). Although these organisms did not persist in the reactor, the continuous addition of these fermenters could decrease the reactor's hydrogen yield. PMID:22156132

  3. Sugarcane molasses-based bio-ethanol wastewater treatment by two-phase multi-staged up-flow anaerobic sludge blanket (UASB) combination with up-flow UASB and down-flow hanging sponge.

    PubMed

    Choeisai, P; Jitkam, N; Silapanoraset, K; Yubolsai, C; Yoochatchaval, W; Yamaguchi, T; Onodera, T; Syutsubo, K

    2014-01-01

    This study was designed to evaluate a treatment system for high strength wastewater (vinasse) from a sugarcane molasses-based bio-ethanol plant in Thailand. A laboratory-scale two-phase treatment system composed of a sulfate reducing (SR) tank and multi-staged up-flow anaerobic sludge blanket (MS-UASB) reactor was used as the pre-treatment unit. Conventional UASB and down-flow hanging sponge (DHS) reactors were used as the post-treatment unit. The treatment system was operated for 300 days under ambient temperature conditions (24.6-29.6 °C). The hydraulic retention time (HRT) in each unit was kept at 25 h for the two-phase system and 23 h for the UASB&DHS. The influent concentration was allowed to reach up to 15,000 mg chemical oxygen demand (COD)/L. COD removal efficiency (based on influent COD) of the two-phase MS-UASB and the UASB&DHS was 54.9 and 18.7%, respectively. Due to the effective removal of sulfide in the SR tank, the MS-UASB achieved a high methane conversion ratio of up to 97%. In DHS, nitrification occurred at the outside portion of the sponge media while denitrification occurred at the inside. Consequently, 27% of the total nitrogen (TN) was removed. An amount of 32% of residual nitrogen (28 mgN/L) was in the form of nitrate, a better nitrogen state for fertilizer. PMID:24647181

  4. Quantitative analysis of previously identified propionate-oxidizing bacteria and methanogens at different temperatures in an UASB reactor containing propionate as a sole carbon source.

    PubMed

    Ban, Qiaoying; Li, Jianzheng; Zhang, Liguo; Jha, Ajay Kumar; Zhang, Yupeng

    2013-12-01

    Propionate degradation is crucial for maintaining the efficiency and stability of an anaerobic reactor. However, there was little information about the effects of ecological factor on propionate-oxidizing bacteria (POB). In current research, quantitative real-time fluorescence polymerase chain reaction (QPCR) of some identified POB and methanogens with a decrease in temperature in an upflow anaerobic sludge bed (UASB) reactor containing propionate as sole carbon source was investigated. The results showed that there were at least four identified POB, including Pelotomaculum schinkii, Pelotomaculum propionicum, Syntrophobacter fumaroxidans, and Syntrophobacter sulfatireducens, observed in this UASB reactor. Among them, P. schinkii was dominated during the whole operational period. Its quantity was 1.2 × 10(4) 16S rRNA gene copies per nanogram of DNA at 35 °C. A decrease in temperature from 35 to 30 °C led to P. schinkii to be increased by 1.8 times and then it was gradually reduced with a decrease in temperature from 30 to 25, 20, and 18 °C stepwise. A decrease in temperature from 35 to 20 °C did not make the amount of methanogens markedly changed, but hydrogenotrophic methanogens (Methanospirillum) and acetotrophic methanogens (Methanosaeta) at 18 °C were increased by an order of magnitude and 1.0 time, respectively, compared with other experimental conditions. PMID:24026412

  5. Assessment of electrochemical and chemical coagulation as post-treatment for the effluents of a UASB reactor treating cellulose pulp mill wastewater.

    PubMed

    Buzzini, A P; Motheo, A J; Pires, E C

    2005-01-01

    This paper presents results from exploratory experiments to test the technical feasibility of electrolytic treatment and coagulation followed by flocculation and sedimentation as post-treatment for the effluent of an UASB reactor treating simulated wastewater from an unbleached Kraft pulp mill. The electrolytic treatment provided up to 67% removal of the remaining COD and 98% of color removal. To achieve these efficiencies the energy consumption ranged from 14 Wh x l(-1) to 20 Wh x l(-1). The coagulation-flocculation treatment followed by settling required 350-400 mg x l(-1) of aluminium sulfate. The addition of a high molecular weight cationic polymer enhanced both COD and color removal. Both post-treatment processes are technically feasible. PMID:16180426

  6. Hydrodynamic characteristics of UASB bioreactors.

    PubMed

    John, Siby; Tare, Vinod

    2011-10-01

    The hydrodynamic characteristics of UASB bioreactors operated under different organic loading and hydraulic loading rates were studied, using three laboratory scale models treating concocted sucrose wastewater. Residence time distribution (RTD) analysis using dispersion model and tanks-in-series model was directed towards the characterization of the fluid flow pattern in the reactors and correlation of the hydraulic regime with the biomass content and biogas production. Empty bed reactors followed a plug flow pattern and the flow pattern changed to a large dispersion mixing with biomass and gas production. Effect of increase in gas production on the overall hydraulics was insignificant. PMID:23505813

  7. The effect of sludge recirculation rate on a UASB-digester treating domestic sewage at 15 °C.

    PubMed

    Zhang, Lei; Hendrickx, Tim L G; Kampman, Christel; Zeeman, Grietje; Temmink, Hardy; Li, Weiguang; Buisman, Cees J N

    2012-01-01

    The anaerobic treatment of low strength domestic sewage at low temperature is an attractive and important topic at present. The upflow anaerobic sludge bed (UASB)-digester system is one of the anaerobic systems to challenge low temperature and concentrations. The effect of sludge recirculation rate on a UASB-digester system treating domestic sewage at 15 °C was studied in this research. A sludge recirculation rate of 0.9, 2.6 and 12.5% of the influent flow rate was investigated. The results showed that the total chemical oxygen demand (COD) removal efficiency rose with increasing sludge recirculation rate. A sludge recirculation rate of 0.9% of the influent flow rate led to organic solids accumulation in the UASB reactor. After the sludge recirculation rate increased from 0.9 to 2.6%, the stability of the UASB sludge was substantially improved from 0.37 to 0.15 g CH?-COD/g COD, and the bio-gas production in the digester went up from 2.9 to 7.4 L/d. The stability of the UASB sludge and bio-gas production in the digester were not significantly further improved by increasing sludge recirculation rate to 12.5% of the influent flow rate, but the biogas production in the UASB increased from 0.37 to 1.2 L/d. It is recommended to apply a maximum sludge recirculation rate of 2-2.5% of the influent flow rate in a UASB-digester system, as this still allows energy self-sufficiency of the system. PMID:23109575

  8. Comparison of Sludge Granule and UASB Performance by Adding Chitosan in Different Forms

    Microsoft Academic Search

    Chantaraporn Phalakornkule

    This study prepared chitosan with the same properties, i.e. molecular weight and degree of deacetylation in different forms, i.e. solution, bead and powder, and compared their effectiveness in enhancing granulation and UASB performance. A natural polymer chitosan in the form of freely moving polymeric chains was found to enhance sludge granulation and the efficiency of UASB system. The UASB with

  9. Operation performance and granule characterization of upflow anaerobic sludge blanket (UASB) reactor treating wastewater with starch as the sole carbon source.

    PubMed

    Lu, Xueqin; Zhen, Guangyin; Estrada, Adriana Ledezma; Chen, Mo; Ni, Jialing; Hojo, Toshimasa; Kubota, Kengo; Li, Yu-You

    2015-03-01

    Long-term performance of a lab-scale UASB reactor treating starch wastewater was investigated under different hydraulic retention times (HRT). Successful start-up could be achieved after 15days' operation. The optimal HRT was 6h with organic loading rate (OLR) 4g COD/Ld at COD concentration 1000mg/L, attaining 81.1-98.7% total COD removal with methane production rate of 0.33L CH4/g CODremoved. Specific methane activity tests demonstrated that methane formation via H2-CO2 and acetate were the principal degradation pathways. Vertical characterizations revealed that main reactions including starch hydrolysis, acidification and methanogenesis occurred at the lower part of reactor ("main reaction zone"); comparatively, at the up converting acetate into methane predominated ("substrate-shortage zone"). Further reducing HRT to 3h caused volatile fatty acids accumulation, sludge floating and performance deterioration. Sludge floating was ascribed to the excess polysaccharides in extracellular polymeric substances (EPS). More efforts are required to overcome sludge floating-related issues. PMID:25617619

  10. Improving hydrolysis of food waste in a leach bed reactor

    SciTech Connect

    Browne, James D.; Allen, Eoin; Murphy, Jerry D., E-mail: jerry.murphy@ucc.ie

    2013-11-15

    Highlights: • This paper assesses leaching of food waste in a two phase digestion system. • Leaching is assessed with and without an upflow anaerobic sludge blanket (UASB). • Without the UASB, low pH reduces hydrolysis, while increased flows increase leaching. • Inclusion of the UASB increases pH to optimal levels and greatly improves leaching. • The optimal conditions are suggested as low flow with connection to the UASB. - Abstract: This paper examines the rate of degradation of food waste in a leach bed reactor (LBR) under four different operating conditions. The effects of leachate recirculation at a low and high flow rate are examined with and without connection to an upflow anaerobic sludge blanket (UASB). Two dilution rates of the effective volume of the leach bed reactors were investigated: 1 and 6 dilutions per LBR per day. The increase in dilution rate from 1 to 6 improved the destruction of volatile solids without connection to the UASB. However connection to the UASB greatly improved the destruction of volatile solids (by almost 60%) at the low recirculation rate of 1 dilution per day. The increase in volatile solids destruction with connection to the UASB was attributed to an increase in leachate pH and buffering capacity provided by recirculated effluent from the UASB to the leach beds. The destruction of volatile solids for both the low and high dilution rates was similar with connection to the UASB, giving 82% and 88% volatile solids destruction respectively. This suggests that the most efficient leaching condition is 1 dilution per day with connection to the UASB.

  11. UASB Treatment of Methanolic Pulp Wastewater with Addition of Waste Starch and Incinerated Ash

    NASA Astrophysics Data System (ADS)

    Takahashi, Shintaro; Kobaysashi, Takuro; Li, Yu-You; Harada, Hideki

    The pulp wastewater consists mainly of methanol. It is expected to treat using upflow anaerobic sludge blanket (UASB) process. Paper manufactories also produce waste starch and incinerated ash. The integrated treating for these wastes is desirable. In this study, two UASB reactors were operated to treat pulp wastewater with addition of waste starch and with addition of incinerated ash, receptively. Continuous operations of a UASB reactor treating pulp wastewater with addition of waste starch (PS reactor) and a UASB reactor treating pulp wastewater with addition of incinerated ash (PA reactor) , were investigated at mesophilic conditions. The PS reactor performed well with an average 93.7% total CODCr and 97.3% soluble CODCr removal efficiency in average at a maximum volumetric loading rate (VLR) of 16.0 kgCOD/m3/d. The PA reactor was also successfully operated with an average 95.3% total CODCr and 97.5% soluble CODCr removal efficiency in average at a maximum VLR of 14.6 kgCOD/m3/d. Successfully developed granules were obtained after over 140 days of operation in both reactors, and the granules were 1 to 2 mm in mean diameter. Microbial analysis revealed the genus Methanomethylovorans was predominant in the granules of both reactors.

  12. Bioelectrochemical enhancement of anaerobic methanogenesis for high organic load rate wastewater treatment in a up-flow anaerobic sludge blanket (UASB) reactor

    PubMed Central

    Zhao, Zhiqiang; Zhang, Yaobin; Chen, Shuo; Quan, Xie; Yu, Qilin

    2014-01-01

    A coupling process of anaerobic methanogenesis and electromethanogenesis was proposed to treat high organic load rate (OLR) wastewater. During the start-up stage, acetate removal efficiency of the electric-biological reactor (R1) reached the maximization about 19 percentage points higher than that of the control anaerobic reactor without electrodes (R2), and CH4 production rate of R1 also increased about 24.9% at the same time, while additional electric input was 1/1.17 of the extra obtained energy from methane. Coulombic efficiency and current recorded showed that anodic oxidation contributed a dominant part in degrading acetate when the metabolism of methanogens was low during the start-up stage. Along with prolonging operating time, aceticlastic methanogenesis gradually replaced anodic oxidation to become the main pathway of degrading acetate. When the methanogens were inhibited under the acidic conditions, anodic oxidation began to become the main pathway of acetate decomposition again, which ensured the reactor to maintain a stable performance. FISH analysis confirmed that the electric field imposed could enrich the H2/H+-utilizing methanogens around the cathode to help for reducing the acidity. This study demonstrated that an anaerobic digester with a pair of electrodes inserted to form a coupling system could enhance methanogenesis and reduce adverse impacts. PMID:25322701

  13. Community onsite treatment of cold strong sewage in a UASB-septic tank.

    PubMed

    Al-Jamal, Wafa; Mahmoud, Nidal

    2009-02-01

    Two community onsite UASB-septic tanks namely R1 and R2 were operated under two different HRT (2 days for R1 and 4 days for R2) in parallel over a year and monitored over the cold half of the year. During the monitoring period, the sewage was characterised by a high COD(tot) of 905mg/l with a high fraction of COD(ss), viz. about 43.7%, and rather low temperature of 17.3 degrees C. The achieved removal efficiencies in R1 and R2 for COD(tot), COD(sus), COD(col), COD(dis), BOD(5) and TSS were "51%, 83%, 20%, 24%, 45% and 74%" and "54%, 87%, 10%, 28%, 49% and 78%", respectively. The difference in the removal efficiencies of those parameters in R1 and R2 is marginal and was only significant (p<0.05) for COD(sus). The sludge filling period of the reactors is expected to be 4-7 years. In view of that, the UASB-septic tank system is a robust and compact system as it can be adequately designed in Palestine at 2 days HRT. PMID:18778934

  14. Improved vortex reactor system

    DOEpatents

    Diebold, James P. (Lakewood, CO); Scahill, John W. (Evergreen, CO)

    1995-01-01

    An improved vortex reactor system for affecting fast pyrolysis of biomass and Refuse Derived Fuel (RDF) feed materials comprising: a vortex reactor having its axis vertically disposed in relation to a jet of a horizontally disposed steam ejector that impels feed materials from a feeder and solids from a recycle loop along with a motive gas into a top part of said reactor.

  15. Reactor water cleanup system

    DOEpatents

    Gluntz, Douglas M. (San Jose, CA); Taft, William E. (Los Gatos, CA)

    1994-01-01

    A reactor water cleanup system includes a reactor pressure vessel containing a reactor core submerged in reactor water. First and second parallel cleanup trains are provided for extracting portions of the reactor water from the pressure vessel, cleaning the extracted water, and returning the cleaned water to the pressure vessel. Each of the cleanup trains includes a heat exchanger for cooling the reactor water, and a cleaner for cleaning the cooled reactor water. A return line is disposed between the cleaner and the pressure vessel for channeling the cleaned water thereto in a first mode of operation. A portion of the cooled water is bypassed around the cleaner during a second mode of operation and returned through the pressure vessel for shutdown cooling.

  16. Reactor water cleanup system

    DOEpatents

    Gluntz, D.M.; Taft, W.E.

    1994-12-20

    A reactor water cleanup system includes a reactor pressure vessel containing a reactor core submerged in reactor water. First and second parallel cleanup trains are provided for extracting portions of the reactor water from the pressure vessel, cleaning the extracted water, and returning the cleaned water to the pressure vessel. Each of the cleanup trains includes a heat exchanger for cooling the reactor water, and a cleaner for cleaning the cooled reactor water. A return line is disposed between the cleaner and the pressure vessel for channeling the cleaned water thereto in a first mode of operation. A portion of the cooled water is bypassed around the cleaner during a second mode of operation and returned through the pressure vessel for shutdown cooling. 1 figure.

  17. Improved vortex reactor system

    DOEpatents

    Diebold, J.P.; Scahill, J.W.

    1995-05-09

    An improved vortex reactor system is described for affecting fast pyrolysis of biomass and Refuse Derived Fuel (RDF) feed materials comprising: a vortex reactor having its axis vertically disposed in relation to a jet of a horizontally disposed steam ejector that impels feed materials from a feeder and solids from a recycle loop along with a motive gas into a top part of said reactor. 12 figs.

  18. Effect of enzymatic pretreatment and increasing the organic loading rate of lipid-rich wastewater treated in a hybrid UASB reactor

    Microsoft Academic Search

    D. R. S. Gomes; L. G. Papa; G. C. V. Cichello; D. Belançon; E. G. Pozzi; J. C. C. Balieiro; E. S. Monterrey-Quintero; G. Tommaso

    2011-01-01

    This study aimed at evaluating the effect of increasing organic loading rates and of enzyme pretreatment on the stability and efficiency of a hybrid upflow anaerobic sludge blanket reactor (UASBh) treating dairy effluent. The UASBh was submitted to the following average organic loading rates (OLR) 0.98Kg.m?3.d?1, 4.58Kg.m?3.d?1, 8.89Kg.m?3.d?1 and 15.73Kg.m?3.d?1, and with the higher value, the reactor was fed with

  19. SCREENING FOR POTENTIAL FERMENTATIVE HYDROGEN PRODUCTION FROM BLACK WATER AND KITCHEN WASTE IN ON?SITE UASB REACTOR AT 20°C

    Microsoft Academic Search

    S. Luostarinen; O. Pakarinen; J. Rintala

    2008-01-01

    The potential of black water and a mixture of black water and kitchen waste as substrates for on?site dark fermentative hydrogen production was screened in upflow anaerobic sludge blanket reactors at 20°C. Three different inocula were used with and without heat treatment. With glucose, the highest specific hydrogenogenic activity was 69 ml H2 g volatile solids d in batch assays

  20. UASB treatment of wastewater containing concentrated benzoate

    SciTech Connect

    Li, Y.Y.; Fang, H.H.P.; Chen, T.; Chui, H.K. [Univ. of Hong Kong (Hong Kong). Civil and Structural Engineering Dept.

    1995-10-01

    The upflow anaerobic sludge blanket (UASB) process removed 97--99% of soluble chemical oxygen demand (COD) from wastewater containing concentrated benzoate at 37 C, pH 7.5, a hydraulic retention time of 9.8 h, and loading rates up to 30.6 g-COD/(L {center_dot} day) based on the reactor volume. About 95.2% of the total COD removed was converted to methane; 0.034 g of volatile suspended solids (VSS) was yielded for each gram of COD removed. The highly settleable granules were 1--3 mm in size with a layered microstructure and were composed in abundance of bacteria resembling the benzoate-degrading Syntrophus buswellii. Two interesting observations have led to the postulation that the degradation of benzoate into acetate was probably conducted completely inside the cell of Syntrophus buswellii-like bacteria: (1) no fatty acids except acetate were found in the effluent; and (2) the granules showed very limited butyrate-degrading capability and could not degrade propionate. This study demonstrated the feasibility of removing aromatic pollutants in wastewater by anaerobic processes.

  1. Quality of brackish aquaculture sludge and its suitability for anaerobic digestion and methane production in an upflow anaerobic sludge blanket (UASB) reactor

    Microsoft Academic Search

    Natella Mirzoyan; Shmuel Parnes; Alon Singer; Yossi Tal; Kevin Sowers; Amit Gross

    2008-01-01

    Intensive recirculating aquaculture systems (RAS) produce high volumes of biosolid waste. The high salinity of brackish\\/marine sludge limits its use in landfill sites and waste outflows and it is a source of pollution. A reduction in sludge mass would therefore minimize the potential environmental hazard and economic burden stemming from its disposal. The aims of the current study were: 1)

  2. Anaerobic treatment of municipal wastewater using the UASB-technology.

    PubMed

    Urban, I; Weichgrebe, D; Rosenwinkel, K-H

    2007-01-01

    The anaerobic treatment of municipal wastewater enables new applications for the reuse of wastewater. The effluent could be used for irrigation as the included nutrients are not affected by the treatment. Much more interesting now are renewable energies and the retrenchment of CO(2) emission. With the anaerobic treatment of municipal wastewater, not only can the CO(2) emission be reduced but "clean" energy supply can be gained by biogas. Most important for the sustainability of this process is the gathering of methane from the liquid effluent of the reactor, because the negative climate-relevant effect from the degassing methane is much higher than the positive effect from saving CO(2) emission. In this study, UASB reactors were used with a flocculent sludge blanket for the biodegradation of the carbon fraction in the wastewater with different temperatures and concentrations. It could be shown that the positive effect is much higher for municipal wastewater with high concentrations in hot climates. PMID:18048975

  3. Nuclear reactor sealing system

    DOEpatents

    McEdwards, James A. (Calabasas, CA)

    1983-01-01

    A liquid metal-cooled nuclear reactor sealing system. The nuclear reactor includes a vessel sealed at its upper end by a closure head. The closure head comprises at least two components, one of which is rotatable; and the two components define an annulus therebetween. The sealing system includes at least a first and second inflatable seal disposed in series in an upper portion of the annulus. The system further includes a dip seal extending into a body of insulation located adjacent a bottom portion of the closure head. The dip seal comprises a trough formed by a lower portion of one of the components, and a seal blade pendently supported from the other component and extending downwardly into the trough. A body of liquid metal is contained in the trough which submerges a portion of the seal blade. The seal blade is provided with at least one aperture located above the body of liquid metal for providing fluid communication between the annulus intermediate the dip seal and the inflatable seals, and a body of cover gas located inside the vessel. There also is provided means for introducing a purge gas into the annulus intermediate the inflatable seals and the seal blade. The purge gas is introduced in an amount sufficient to substantially reduce diffusion of radioactive cover gas or sodium vapor up to the inflatable seals. The purge gas mixes with the cover gas in the reactor vessel where it can be withdrawn from the vessel for treatment and recycle to the vessel.

  4. Nuclear reactor shutdown system

    DOEpatents

    Bhate, Suresh K. (Niskayuna, NY); Cooper, Martin H. (Monroeville, PA); Riffe, Delmar R. (Murrysville, PA); Kinney, Calvin L. (Penn Hills, PA)

    1981-01-01

    An inherent shutdown system for a nuclear reactor having neutron absorbing rods affixed to an armature which is held in an upper position by a magnetic flux flowing through a Curie temperature material. The Curie temperature material is fixedly positioned about the exterior of an inner duct in an annular region through which reactor coolant flows. Elongated fuel rods extending from within the core upwardly toward the Curie temperature material are preferably disposed within the annular region. Upon abnormal conditions which result in high neutron flux and coolant temperature, the Curie material loses its magnetic permeability, breaking the magnetic flux path and allowing the armature and absorber rods to drop into the core, thus shutting down the fissioning reaction. The armature and absorber rods are retrieved by lowering the housing for the electromagnet forming coils which create a magnetic flux path which includes the inner duct wall. The coil housing then is raised, resetting the armature.

  5. Attrition reactor system

    DOEpatents

    Scott, Charles D. (Oak Ridge, TN); Davison, Brian H. (Knoxvile, TN)

    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.

  6. Attrition reactor system

    DOEpatents

    Scott, C.D.; Davison, B.H.

    1993-09-28

    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. 2 figures.

  7. Anaerobic/aerobic treatment of greywater via UASB and MBR for unrestricted reuse.

    PubMed

    Abdel-Shafy, Hussein I; Al-Sulaiman, Ahmed Makki; Mansour, Mona S M

    2015-01-01

    The aim of the present study was to investigate the efficiency of integrated up-flow anaerobic sludge blanket (UASB) as anaerobic system followed by membrane bioreactor (MBR) as aerobic system for the treatment of greywater for unrestricted reuse. Pilot-scale UASB and MBR units were installed and operated in the NRC, Egypt. Real raw greywater was subjected to UASB and the effluent was further treated with microfiltration MBR. The necessary trans-membrane pressure difference is applied by the water head above the membrane (gravity flow) without any energy input. The average characteristics of the raw greywater were 95, 392, 298, 10.45, 0.4, 118.5 and 28 mg/L for total suspended solids (TSS), chemical oxygen demand (COD), biochemical oxygen demand (BOD), total phosphates, nitrates, oil and grease, and total Kjeldahl nitrogen (TKN), respectively. The pH was 6.71. The UASB treatment efficiency reached 19.3, 57.8, 67.5 and 83.7% for TSS, COD, BOD5 and oil and grease, respectively. When the UASB effluent was further treated with MBR, the overall removal rate achieved 97.7, 97.8, 97.4 and 95.8% for the same parameters successively. The characteristics of the final effluent reached 2.5, 8.5, 6.1, 0.95, 4.6 and 2.3 mg/L for TSS, COD, BOD, phosphates, oil and grease and TKN, respectively. This final treated effluent could cope with the unrestricted water reuse of local Egyptian guidelines. PMID:25746657

  8. Reactor vessel support system. [LMFBR

    DOEpatents

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

    1980-05-09

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

  9. The role of sludge retention time in the hydrolysis and acidification of lipids, carbohydrates and proteins during digestion of primary sludge in CSTR systems

    Microsoft Academic Search

    Yehuda Miron; Grietje Zeeman; Jules B van Lier; Gatze Lettinga

    2000-01-01

    The effect of the sludge retention time (SRT) between 3 and 15days, on hydrolysis, acidification and methanogenesis of domestic sewage was researched by simulating a sludge bed segment of an upflow anaerobic sludge bed (UASB) system as a completely stirred tank reactor (CSTR). The CSTR systems were fed with primary sludge (settled solids of domestic sewage) as the influent at

  10. Removal of Total Coliforms, Thermotolerant Coliforms, and Helminth Eggs in Swine Production Wastewater Treated in Anaerobic and Aerobic Reactors

    PubMed Central

    Zacarias Sylvestre, Silvia Helena; Lux Hoppe, Estevam Guilherme; de Oliveira, Roberto Alves

    2014-01-01

    The present work evaluated the performance of two treatment systems in reducing indicators of biological contamination in swine production wastewater. System I consisted of two upflow anaerobic sludge blanket (UASB) reactors, with 510 and 209?L in volume, being serially arranged. System II consisted of a UASB reactor, anaerobic filter, trickling filter, and decanter, being also organized in series, with volumes of 300, 190, 250, and 150?L, respectively. Hydraulic retention times (HRT) applied in the first UASB reactors were 40, 30, 20, and 11?h in systems I and II. The average removal efficiencies of total and thermotolerant coliforms in system I were 92.92% to 99.50% and 94.29% to 99.56%, respectively, and increased in system II to 99.45% to 99.91% and 99.52% to 99.93%, respectively. Average removal rates of helminth eggs in system I were 96.44% to 99.11%, reaching 100% as in system II. In reactor sludge, the counts of total and thermotolerant coliforms ranged between 105 and 109?MPN (100?mL)?1, while helminth eggs ranged from 0.86 to 9.27?eggs?g?1?TS. PMID:24812560

  11. A case study of enteric virus removal and insights into the associated risk of water reuse for two wastewater treatment pond systems in Bolivia.

    PubMed

    Symonds, E M; Verbyla, M E; Lukasik, J O; Kafle, R C; Breitbart, M; Mihelcic, J R

    2014-11-15

    Wastewater treatment ponds (WTP) are one of the most widespread treatment technologies in the world; however, the mechanisms and extent of enteric virus removal in these systems are poorly understood. Two WTP systems in Bolivia, with similar overall hydraulic retention times but different first stages of treatment, were analyzed for enteric virus removal. One system consisted of a facultative pond followed by two maturation ponds (three-pond system) and the other consisted of an upflow anaerobic sludge blanket (UASB) reactor followed by two maturation (polishing) ponds (UASB-pond system). Quantitative polymerase chain reaction with reverse transcription (RT-qPCR) was used to measure concentrations of norovirus, rotavirus, and pepper mild mottle virus, while cell culture methods were used to measure concentrations of culturable enteroviruses (EV). Limited virus removal was observed with RT-qPCR in either system; however, the three-pond system removed culturable EV with greater efficiency than the UASB-pond system. The majority of viruses were not associated with particles and only a small proportion was associated with particles larger than 180 ?m; thus, it is unlikely that sedimentation is a major mechanism of virus removal. High concentrations of viruses were associated with particles between 0.45 and 180 ?m in the UASB reactor effluent, but not in the facultative pond effluent. The association of viruses with this size class of particles may explain why only minimal virus removal was observed in the UASB-pond system. Quantitative microbial risk assessment of the treated effluent for reuse for restricted irrigation indicated that the three-pond system effluent requires an additional 1- to 2-log10 reduction of viruses to achieve the WHO health target of <10(-4) disability-adjusted life years (DALYs) lost per person per year; however, the UASB-pond system effluent may require an additional 2.5- to 4.5-log10 reduction of viruses. PMID:25129566

  12. Decolorization and COD reduction of UASB pretreated poultry manure wastewater by electrocoagulation process: A post-treatment study

    Microsoft Academic Search

    Kaan Yetilmezsoy; Fatih Ilhan; Zehra Sapci-Zengin; Suleyman Sakar; M. Talha Gonullu

    2009-01-01

    The performance of electrocoagulation (EC) technique for decolorization and chemical oxygen demand (COD) reduction of anaerobically pretreated poultry manure wastewater was investigated in a laboratory batch study. Two identical 15.7-L up-flow anaerobic sludge blanket (UASB) reactors were first run under various organic and hydraulic loading conditions for 216 days. Effects of operating parameters such as type of sacrificial electrode material,

  13. Influence of hydraulic retention time on fouling in a UASB coupled with an external ultrafiltration membrane treating synthetic municipal wastewater

    Microsoft Academic Search

    Mónica L. Salazar-Peláez; Juan M. Morgan-Sagastume; Adalberto Noyola

    2011-01-01

    This paper evaluates the effect of hydraulic retention time (HRT)11HRT: hydraulic retention time. on anaerobic membrane bioreactor (AnMBR)22AnMBR: anaerobic membrane bioreactor. performance, biopolymeric substances production and fouling; also, struvite precipitation is analyzed based on solubility product equation. A lab-scale UASB reactor coupled with an ultrafiltration (UF)33UF: ultrafiltration. membrane was operated under three different HRTs (4, 8 and 12h) and fed

  14. Membrane installation for enhanced up-flow anaerobic sludge blanket (UASB) performance.

    PubMed

    Liu, Yin; Zhang, Kaisong; Bakke, Rune; Li, Chunming; Liu, Haining

    2013-09-01

    It is postulated that up-flow anaerobic sludge blanket (UASB) reactor efficiency can be enhanced by a membrane immersed in the reactor to operate it as an anaerobic membrane bioreactor (AnMBR) for low-strength wastewater treatment. This postulate was tested by comparing the performance with and without a hollow fiber microfiltration membrane module immersed in UASB reactors operated at two specific organic loading rates (SOLR). Results showed that membrane filtration enhanced process performance and stability, with over 90% total organic carbon (TOC) removal consistently achieved. More than 91% of the TOC removal was achieved by suspended biomass, while less than 6% was removed by membrane filtration and digestion in the membrane attached biofilm during stable AnMBRs operation. Although the membrane and its biofilm played an important role in initial stage of the high SOLR test, linear increased TOC removal by bulk sludge mainly accounted for the enhanced process performance, implying that membrane led to enhanced biological activity of the suspended sludge. The high retention of active fine sludge particles in suspension was the main reason for this significant improvement of performance and biological activity, which led to decreased SOLR with time to a theoretical optimal level around 2  g COD/g MLVSS·d and the establishment of a microbial community dominated by Methanothrix-like microbes. It was concluded that UASB process performance can be enhanced by transforming such to AnMBR operation when the loading rate is too high for sufficient sludge retention, and/or when the effluent water quality demands are especially stringent. PMID:23578587

  15. Reactor core isolation cooling system

    DOEpatents

    Cooke, F.E.

    1992-12-08

    A reactor core isolation cooling system includes a reactor pressure vessel containing a reactor core, a drywell vessel, a containment vessel, and an isolation pool containing an isolation condenser. A turbine is operatively joined to the pressure vessel outlet steamline and powers a pump operatively joined to the pressure vessel feedwater line. In operation, steam from the pressure vessel powers the turbine which in turn powers the pump to pump makeup water from a pool to the feedwater line into the pressure vessel for maintaining water level over the reactor core. Steam discharged from the turbine is channeled to the isolation condenser and is condensed therein. The resulting heat is discharged into the isolation pool and vented to the atmosphere outside the containment vessel for removing heat therefrom. 1 figure.

  16. Reactor core isolation cooling system

    DOEpatents

    Cooke, Franklin E. (San Jose, CA)

    1992-01-01

    A reactor core isolation cooling system includes a reactor pressure vessel containing a reactor core, a drywell vessel, a containment vessel, and an isolation pool containing an isolation condenser. A turbine is operatively joined to the pressure vessel outlet steamline and powers a pump operatively joined to the pressure vessel feedwater line. In operation, steam from the pressure vessel powers the turbine which in turn powers the pump to pump makeup water from a pool to the feedwater line into the pressure vessel for maintaining water level over the reactor core. Steam discharged from the turbine is channeled to the isolation condenser and is condensed therein. The resulting heat is discharged into the isolation pool and vented to the atmosphere outside the containment vessel for removing heat therefrom.

  17. Reactor refueling containment system

    DOEpatents

    Gillett, J.E.; Meuschke, R.E.

    1995-05-02

    A method of refueling a nuclear reactor is disclosed whereby the drive mechanism is disengaged and removed by activating a jacking mechanism that raises the closure head. The area between the barrier plate and closure head is exhausted through the closure head penetrations. The closure head, upper drive mechanism, and bellows seal are lifted away and transported to a safe area. The barrier plate acts as the primary boundary and each drive and control rod penetration has an elastomer seal preventing excessive tritium gases from escaping. The individual instrumentation plugs are disengaged allowing the corresponding fuel assembly to be sealed and replaced. 2 figs.

  18. Reactor refueling containment system

    SciTech Connect

    Gillett, J.E.; Meuschke, R.E.

    1992-12-31

    This report describes a method of refueling a nuclear reactor whereby the drive mechanism is disengaged and removed by activating a jacking mechanism that raises the closure head. The area between the barrier plate and closure head is exhausted through the closure head penetrations. The closure head, upper drive mechanism, and bellows seal are lifted away and transported to a safe area. The barrier plate acts as the primary boundary and each drive and control rod penetration has an elastomer seal preventing excessive tritium gases from escaping. The individual instrumentation plugs are disengaged allowing the corresponding fuel assembly to be sealed and replaced.

  19. Reactor refueling containment system

    DOEpatents

    Gillett, James E. (Greensburg, PA); Meuschke, Robert E. (Pittsburgh, PA)

    1995-01-01

    A method of refueling a nuclear reactor whereby the drive mechanism is disengaged and removed by activating a jacking mechanism that raises the closure head. The area between the barrier plate and closure head is exhausted through the closure head penetrations. The closure head, upper drive mechanism, and bellows seal are lifted away and transported to a safe area. The barrier plate acts as the primary boundary and each drive and control rod penetration has an elastomer seal preventing excessive tritium gases from escaping. The individual instrumentation plugs are disengaged allowing the corresponding fuel assembly to be sealed and replaced.

  20. Plasma reactor waste management systems

    NASA Technical Reports Server (NTRS)

    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.

  1. Solvent refined coal reactor quench system

    DOEpatents

    Thorogood, Robert M. (Macungie, PA)

    1983-01-01

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

  2. Solvent refined coal reactor quench system

    DOEpatents

    Thorogood, R.M.

    1983-11-08

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

  3. Rapid starting methanol reactor system

    DOEpatents

    Chludzinski, Paul J. (38 Berkshire St., Swampscott, MA 01907); Dantowitz, Philip (39 Nancy Ave., Peabody, MA 01960); McElroy, James F. (12 Old Cart Rd., Hamilton, MA 01936)

    1984-01-01

    The invention relates to a methanol-to-hydrogen cracking reactor for use with a fuel cell vehicular power plant. The system is particularly designed for rapid start-up of the catalytic methanol cracking reactor after an extended shut-down period, i.e., after the vehicular fuel cell power plant has been inoperative overnight. Rapid system start-up is accomplished by a combination of direct and indirect heating of the cracking catalyst. Initially, liquid methanol is burned with a stoichiometric or slightly lean air mixture in the combustion chamber of the reactor assembly. The hot combustion gas travels down a flue gas chamber in heat exchange relationship with the catalytic cracking chamber transferring heat across the catalyst chamber wall to heat the catalyst indirectly. The combustion gas is then diverted back through the catalyst bed to heat the catalyst pellets directly. When the cracking reactor temperature reaches operating temperature, methanol combustion is stopped and a hot gas valve is switched to route the flue gas overboard, with methanol being fed directly to the catalytic cracking reactor. Thereafter, the burner operates on excess hydrogen from the fuel cells.

  4. UASB performance and microbial adaptation during a transition from mesophilic to thermophilic treatment of palm oil mill effluent.

    PubMed

    Khemkhao, Maneerat; Nuntakumjorn, Boonyarit; Techkarnjanaruk, Somkiet; Phalakornkule, Chantaraporn

    2012-07-30

    The treatment of palm oil mill effluent (POME) by an upflow anaerobic sludge bed (UASB) at organic loading rates (OLR) between 2.2 and 9.5 g COD l(-1) day(-1) was achieved by acclimatizing the mesophilic (37 °C) microbial seed to the thermophilic temperature (57 °C) by a series of stepwise temperature shifts. The UASB produced up to 13.2 l biogas d(-1) with methane content on an average of 76%. The COD removal efficiency ranged between 76 and 86%. Microbial diversity of granules from the UASB reactor was also investigated. The PCR-based DGGE analysis showed that the bacterial population profiles significantly changed with the temperature transition from mesophilic to thermophilic conditions. In addition, the results suggested that even though the thermophilic temperature of 57 °C was suitable for a number of hydrolytic, acidogenic and acetogenic bacteria, it may not be suitable for some Methanosaeta species acclimatized from 37 °C. Specifically, the bands associated with Methanosaeta thermophila PT and Methanosaeta harundinacea can be detected during the four consecutive operation phases of 37 °C, 42 °C, 47 °C and 52 °C, but their corresponding bands were found to fade out at 57 °C. The DGGE analysis predicted that the temperature transition can result in significant methanogenic biomass washout at 57 °C. PMID:22466006

  5. Acetate degradation at 70[degrees]C in upflow anaerobic sludge blanket reactors and temperature response of granules grown at 70[degrees]C

    SciTech Connect

    Rintala, J.; Lepistoe, S. (Tampere Univ. of Technology (Finland)); Ahring, B. (Technical Univ., Lyngby (Denmark))

    1993-06-01

    Several industrial production processes generate hot, concentrated process water and wastewater streams. Recent studies have demonstrated that hot industrial wastewaters can also profitably be treated anaerobically. Sludge granulation occurs at 55[degrees]C in thermophilic UASB reactors and the UASB process has been demonstrated feasible for industrial wastewater treatment at 55[degrees]C. However, very little is known about anaerobic wastewater treatment at temperatures above 55[degrees]C. This study investigates acetate degradation in UASB reactors at 70[degrees]C and characterizes the temperature response of the reactor sludges compared with sludge from a UASB reactor at 55[degrees]C. Results show that high concentrations of acetate can be efficiently converted to methane even at 70[degrees]C. However, the start-up of the UASB reactors was slower and the specific sludge activities were lower at 70[degrees]C than at 55[degrees]C. 21 refs., 3 figs., 2 tabs.

  6. Liquid metal cooled nuclear reactor plant system

    DOEpatents

    Hunsbedt, Anstein (Los Gatos, CA); Boardman, Charles E. (Saratoga, CA)

    1993-01-01

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

  7. Shutdown system for a nuclear reactor

    DOEpatents

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

    1984-06-05

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

  8. Shutdown system for a nuclear reactor

    DOEpatents

    Groh, Edward F. (Naperville, IL); Olson, Arne P. (Western Springs, IL); Wade, David C. (Naperville, IL); Robinson, Bryan W. (Oak Lawn, IL)

    1984-01-01

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

  9. EFFECT OF STARCH ADDITION ON THE PERFORMANCE AND SLUDGE CHARACTERIZATION OF UASB PROCESS TREATING METHANOLIC WASTEWATER

    NASA Astrophysics Data System (ADS)

    Yan, Feng; Kobayashi, Takuro; Takahashi, Shintaro; Li, Yu-You; Omura, Tatsuo

    A mesophilic(35℃) UASB reactor treating synthetic wastewater containing methanol with addition of starch was continuously operated for over 430 days by changing the organic loading rate from 2.5 to 120kg-COD/m3.d. The microbial community structure of the granules was analyzed with the molecular tools and its metabolic characteristics were evaluated using specific methanogenic activity tests. The process was successfully operated with over 98% soluble COD removal efficiency at VLR 30kg-COD/m3.d for approximately 300 days, and granulation satisfactory proceeded. The results of cloning and fluorescence in situ hybridization analysis suggest that groups related the genus Methanomethylovorans and the genus Methanosaeta were predominant in the reactor although only the genus Methanomethylovorans was predominant in the reactor treating methanolic wastewater in the previous study. Abundance of the granules over 0.5 mm in diameter in the reactor treating methanolic wastewater with addition of starch was 3 times larger than that in the reactor treating methanolic wastewater. Specific methanogenic activity tests in this study indicate that the methanol-methane pathway and the methanol-H2/CO2-methane pathway were predominant, and however, there was a certain level of activity for acetate-methane pathway unlike the reactor treating methanolic wastewater. These results suggest addition of starch might be responsible for diversifying the microbial community and encouraging the granulation.

  10. Nuclear electric propulsion reactor control systems status

    NASA Technical Reports Server (NTRS)

    Ferg, D. A.

    1973-01-01

    The thermionic reactor control system design studies conducted over the past several years for a nuclear electric propulsion system are described and summarized. The relevant reactor control system studies are discussed in qualitative terms, pointing out the significant advantages and disadvantages including the impact that the various control systems would have on the nuclear electric propulsion system design. A recommendation for the reference control system is made, and a program for future work leading to an engineering model is described.

  11. Reductive decolourisation of sulphonated mono and diazo dyes in one- and two-stage anaerobic systems.

    PubMed

    da Silva, Marcos Erick Rodrigues; Firmino, Paulo Igor Milen; dos Santos, André Bezerra

    2013-05-01

    This work assessed the application of one- and two-stage mesophilic anaerobic systems to colour removal of sulphonated mono and diazo dyes with ethanol as electron donor. The dyes Congo Red (CR), Reactive Black 5 (RB5) and Reactive Red 2 (RR2) were selected as model compounds and tested separately in seven different periods. The one-stage system (R(1)) consisted of a single up-flow anaerobic sludge blanket (UASB) reactor, whereas the two-stage system (R(2)) consisted of an acidogenic UASB reactor (R(A)), a settler and a methanogenic UASB reactor (R(M)). For CR and RB5, no remarkable difference was observed between the colour removal performance of both anaerobic systems R(1) and R(2). The experiments with RR2 revealed that R(2) was more efficient on colour removal than R(1), showing efficiencies almost 2-fold (period VI) and 2.5-fold (period VII) higher than those found by R(1). Additionally, R(2) showed a higher stability, giving a good prospect for application to textile wastewaters. Finally, the acidogenic reactor (R(A)) had an important role in the overall decolourisation achieved by R(2) during the experiments with CR and RB5 (>78 %), whereas for RR2, a more recalcitrant dye, R(A) was responsible for up to 38 % of the total colour removal. PMID:23456307

  12. The 5-kwe reactor thermoelectric system summary

    NASA Technical Reports Server (NTRS)

    Vanosdol, J. H. (editor)

    1973-01-01

    Design of the 5-kwe reactor thermoelectric system was initiated in February 1972 and extended through the conceptual design phase into the preliminary design phase. Design effort was terminated in January, 1973. This report documents the system and component requirements, design approaches, and performance and design characteristics for the 5-kwe system. Included is summary information on the reactor, radiation shields, power conversion systems, thermoelectric pump, radiator/structure, liquid metal components, and the control system.

  13. Small reactor power system for space application

    NASA Technical Reports Server (NTRS)

    Shirbacheh, M.

    1987-01-01

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

  14. Anaerobic on-site treatment of black water and dairy parlour wastewater in UASB-septic tanks at low temperatures.

    PubMed

    Luostarinen, Sari A; Rintala, Jukka A

    2005-01-01

    Anaerobic on-site treatment of synthetic black water (BW) and dairy parlour wastewater (DPWW) was studied in two-phased upflow anaerobic sludge blanket (UASB)-septic tanks at low temperatures (10-20 degrees C). At all temperatures, total chemical oxygen demand (COD(t)) removal was above 90% with BW and above 80% with DPWW and removal of total suspended solids (TSS) above 90% with both wastewaters. Moreover, dissolved COD (COD(dis)) removal was approx. 70% with both wastewaters indicating good biological activity of the sludges. With BW, a single-phased reactor was found sufficient for good COD removals, while with DPWW, a two-phased process was required. Temperature optimum of reactor sludges was still 35 degrees C after long (398d) operation. Most of the nutrients from BW were removed with TSS, while with DPWW nutrient removal was low. In conclusion, UASB-septic tank was found feasible for (pre)treatment of BW and DPWW at low temperatures. PMID:15644252

  15. TREAT Reactor Control and Protection System

    SciTech Connect

    Lipinski, W.C.; Brookshier, W.K.; Burrows, D.R.; Lenkszus, F.R.; McDowell, W.P.

    1985-01-01

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

  16. Self-actuating reactor shutdown system

    DOEpatents

    Barrus, Donald M. (San Jose, CA); Brummond, Willian A (Livermore, CA); Peterson, Leslie F. (Danville, CA)

    1988-01-01

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

  17. Reactor power system/spacecraft integration

    NASA Technical Reports Server (NTRS)

    Elms, R. V.

    1985-01-01

    The new national initiative in space reactor technology evaluation and development is strongly tied to mission applications and to spacecraft and space transportation system (STS) compatibility. This paper discusses the power system integration interfaces with potential using spacecraft and the STS, and the impact of these requirements on the design. The integration areas of interest are mechanical, thermal, electrical, attitude control, and mission environments. The mission environments include space vacuum, solar input, heat sink, space radiation, weapons effects, and reactor power system radiation environments. The natural, reactor, and weapons effects radiation must be evaluated and combined to define the design requirements for spacecraft electronic equipment.

  18. REACTOR: An Expert System for Diagnosis and Treatment of Nuclear Reactor Accidents

    Microsoft Academic Search

    William R. Nelson

    1982-01-01

    REACTOR is an expert system under development at EG&G Idaho, Inc., that will assist operators in the diagnosis and treatment of nuclear reactor accidents. This paper covers the background of the nuclear industry and why expert system tech- nology may prove valuable in the reactor control room. Some of the basic features of the REACTOR system are discussed, and future

  19. Physical modelling of an upflow anaerobic sludge blanket reactor: near-field study.

    PubMed

    Gimenez, J R; Nassr, S C; Maestri, R D; Monteggla, L O

    2002-01-01

    This paper presents a physical evaluation of an upflow anaerobic sludge blanket reactor. Specifically, the study contemplates the region influenced by the wastewater inlet jets at the bottom of the reactor, here termed the near-field area. A three-dimensional physical model of a UASB reactor in reduced scale, geometrically and dynamically correlated to a full-scale prototype was used in the evaluation. From the analysis of the major forces acting and of the physical properties investigated in the prototype, it was possible to establish non-dimensional relations that were applied to the reduced scale model, allowing the investigation of the phenomenon of interest. Tests were developed on the model to visualise the inlet flows under the buoyant effect at the bottom of the reactor, through the injection of a tracer fluid in the area with higher density, simulating the effects of the sludge bed. Based on the experimental results, it was possible to provide dimensioning criteria for the jet distribution system in UASB reactors. PMID:12188537

  20. Scanning tunneling microscope assembly, reactor, and system

    SciTech Connect

    Tao, Feng; Salmeron, Miquel; Somorjai, Gabor A

    2014-11-18

    An embodiment of a scanning tunneling microscope (STM) reactor includes a pressure vessel, an STM assembly, and three spring coupling objects. The pressure vessel includes a sealable port, an interior, and an exterior. An embodiment of an STM system includes a vacuum chamber, an STM reactor, and three springs. The three springs couple the STM reactor to the vacuum chamber and are operable to suspend the scanning tunneling microscope reactor within the interior of the vacuum chamber during operation of the STM reactor. An embodiment of an STM assembly includes a coarse displacement arrangement, a piezoelectric fine displacement scanning tube coupled to the coarse displacement arrangement, and a receiver. The piezoelectric fine displacement scanning tube is coupled to the coarse displacement arrangement. The receiver is coupled to the piezoelectric scanning tube and is operable to receive a tip holder, and the tip holder is operable to receive a tip.

  1. Flexible Conversion Ratio Fast Reactor Systems Evaluation

    SciTech Connect

    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.

  2. 10 CFR 50.46a - Acceptance criteria for reactor coolant system venting systems.

    Code of Federal Regulations, 2013 CFR

    2013-01-01

    ...Regulatory Approvals § 50.46a Acceptance criteria for reactor coolant system venting systems. Each nuclear power reactor must be provided with high point vents for the reactor coolant system, for the reactor vessel head, and for...

  3. 10 CFR 50.46a - Acceptance criteria for reactor coolant system venting systems.

    Code of Federal Regulations, 2014 CFR

    2014-01-01

    ...Regulatory Approvals § 50.46a Acceptance criteria for reactor coolant system venting systems. Each nuclear power reactor must be provided with high point vents for the reactor coolant system, for the reactor vessel head, and for...

  4. 10 CFR 50.46a - Acceptance criteria for reactor coolant system venting systems.

    Code of Federal Regulations, 2012 CFR

    2012-01-01

    ...Regulatory Approvals § 50.46a Acceptance criteria for reactor coolant system venting systems. Each nuclear power reactor must be provided with high point vents for the reactor coolant system, for the reactor vessel head, and for...

  5. Performance and microbial community analysis of a pilot-scale UASB for corn-ethanol wastewater treatment.

    PubMed

    Huang, Jianping; Xiao, Ling; Xi, Chunhui

    2015-04-01

    The performance and microbial community structure of a pilot-scale upflow anaerobic sludge blanket (UASB) reactor inoculated with flocculent sludge were investigated over 52 days. The characteristics of corn-ethanol wastewater were as follows: CODCr, 1,050-4,970 mg l(-1); ammonia, 14-298 mg l(-1); and alkalinity, 332-2,867 mg l(-1). The UASB could start up smoothly with a hydraulic loading rate lower than 180 l h(-1) and a ratio of volatile fatty acid versus alkalinity between 0.04 and 0.48. The maximum gas production rate was 432 l h(-1) and the highest volumetric loading rate of 7.2 kg m(-3) day(-1) was obtained after 48 days. The 1 mm granules could form a complex network and were composed of many Methanosaeta. Aceticlastic methanogens served as a dominant methanogenic group, which accounted for the relatively high resistance to shock loading. PMID:25537339

  6. Gaseous fuel reactors for power systems

    NASA Technical Reports Server (NTRS)

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

    1977-01-01

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

  7. Control system for a small fission reactor

    DOEpatents

    Burelbach, James P. (Glen Ellyn, IL); Kann, William J. (Park Ridge, IL); Saiveau, James G. (Hickory Hills, IL)

    1986-01-01

    A system for controlling the reactivity of a small fission reactor includes an elongated, flexible hollow tube in the general form of a helical coiled spring axially positioned around and outside of the reactor vessel in an annular space between the reactor vessel and a surrounding cylindrical-shaped neutron reflector. A neutron absorbing material is provided within the hollow tube with the rate of the reaction controlled by the extension and compression of the hollow tube, e.g., extension of the tube increases reactivity while its compression reduces reactivity, in varying the amount of neutron absorbing material disposed between the reactor vessel and the neutron reflector. Conventional mechanical displacement means may be employed to control the coil density of the hollow tube as desired. In another embodiment, a plurality of flexible hollow tubes each containing a neutron absorber are positioned adjacent to one another in spaced relation around the periphery of the reactor vessel and inside the outer neutron reflector with reactivity controlled by the extension and compression of all or some of the coiled hollow tubes. Yet another embodiment of the invention envisions the neutron reflector in the form of an expandable coil spring positioned in an annular space between the reactor vessel and an outer neutron absorbing structure for controlling the neutron flux reflected back into the reactor vessel.

  8. Weld monitor and failure detector for nuclear reactor system

    Microsoft Academic Search

    H. G. Jr

    1987-01-01

    This patent describes a detection system for monitoring the integrity of inaccessible weld means throughout the life of a reactor system and providing a responsive signal indicative of weld failure. The nuclear reactor system includes a steel nuclear reactor vessel having an outer wall surface and an inner wall surface and reactor vessel structural support means projecting inwardly from the

  9. [Effect of pilot UASB-SFSBR-MAP process for the large scale swine wastewater treatment].

    PubMed

    Wang, Liang; Chen, Chong-Jun; Chen, Ying-Xu; Wu, Wei-Xiang

    2013-03-01

    In this paper, a treatment process consisted of UASB, step-fed sequencing batch reactor (SFSBR) and magnesium ammonium phosphate precipitation reactor (MAP) was built to treat the large scale swine wastewater, which aimed at overcoming drawbacks of conventional anaerobic-aerobic treatment process and SBR treatment process, such as the low denitrification efficiency, high operating costs and high nutrient losses and so on. Based on the treatment process, a pilot engineering was constructed. It was concluded from the experiment results that the removal efficiency of COD, NH4(+) -N and TP reached 95.1%, 92.7% and 88.8%, the recovery rate of NH4(+) -N and TP by MAP process reached 23.9% and 83.8%, the effluent quality was superior to the discharge standard of pollutants for livestock and poultry breeding (GB 18596-2001), mass concentration of COD, TN, NH4(+) -N, TP and SS were not higher than 135, 116, 43, 7.3 and 50 mg x L(-1) respectively. The process developed was reliable, kept self-balance of carbon source and alkalinity, reached high nutrient recovery efficiency. And the operating cost was equal to that of the traditional anaerobic-aerobic treatment process. So the treatment process could provide a high value of application and dissemination and be fit for the treatment pf the large scale swine wastewater in China. PMID:23745404

  10. Effect of chitosan on UASB treating POME during a transition from mesophilic to thermophilic conditions.

    PubMed

    Khemkhao, Maneerat; Nuntakumjorn, Boonyarit; Techkarnjanaruk, Somkiet; Phalakornkule, Chantaraporn

    2011-04-01

    The effects of chitosan addition on treatment of palm oil mill effluent were investigated using two lab-scale upflow anaerobic sludge bed (UASB) reactors: (1) with chitosan addition at the dosage of 2 mg chitosan per g volatile suspended solids on the first day of the operation (R1), (2) without chitosan addition (the control, R2). The reactors were inoculated with mesophilic anaerobic sludge which was acclimatized to a thermophilic condition with a stepwise temperature increase of 5 °C from 37 to 57 °C. The OLR ranged from 2.23 to 9.47 kg COD m(-3) day(-1). The difference in biogas production rate increased from non-significant to 18% different. The effluent volatile suspended solids of R1 was 65 mg l(-1) lower than that of R2 on Day 123. 16S rRNA targeted denaturing gradient gel electrophoresis (DGGE) fingerprints of microbial community indicated that some methanogens in the genus Methanosaeta can be detected in R1 but not in R2. PMID:21316949

  11. Space reactor power system programs overview

    SciTech Connect

    Bloomfield, H.S. (NASA, Lewis Research Center, Cleveland, OH (United States))

    1992-03-01

    The present development history and current development status evaluation of space reactor power system technologies gives attention to subsystem and component readiness and performance, and assesses the technology data base available in each case. This data base characterization gives attention to the most compatible reactor-power conversion system combinations for prospective DOD and commercial missions, as well as NASA missions. Candidate systems for near, middle, and far term application are selected and prioritized on the basis of technical risk. The programs covered encompass SNAPs 1, 2, 8, and 10A, SNAP 50, and SP-100. 6 refs.

  12. Control system for a small fission reactor

    DOEpatents

    Burelbach, J.P.; Kann, W.J.; Saiveau, J.G.

    1985-02-08

    A system for controlling the reactivity of a small fission reactor includes an elongated, flexible hollow tube in the general form of a helical coiled spring axially positioned around and outside of the reactor vessel in an annular space between the reactor vessel and a surrounding cylindrical-shaped neutron reflector. A neutron absorbing material is provided within the hollow tube with the rate of the reaction controlled by the extension and compression of the hollow tube, e.g., extension of the tube increases reactivity while its compression reduces reactivity, in varying the amount of neutron absorbing material disposed between the reactor vessel and the neutron reflector. Conventional mechanical displacement means may be employed to control the coil density of the hollow tube as desired.

  13. Liquid lithium divertor system for fusion reactor

    Microsoft Academic Search

    Yoshio Nagayama

    2009-01-01

    One of the most critical issues for the steady state fusion reactor is the heat flux in the divertor target. This paper proposes a liquid lithium divertor system to solve this problem. The proposed divertor system consists of a liquid lithium target, an evaporation chamber and a differential evacuation chamber. The heat coming from the fusion plasma along the divertor

  14. Systems Issues in Nuclear Reactor Safety

    E-print Network

    de Weck, Olivier L.

    Systems Issues in Nuclear Reactor Safety Commissioner George ApostolakisCommissioner George Apostolakis U.S. Nuclear Regulatory Commission CmrApostolakis@nrc.gov MIT SDM Conference on Systems Thinking, source, and special nuclear materials to ensure adequate protection of public health and safety, 3

  15. Anaerobic digestion of ice-cream wastewater: A comparison of single and two-phase reactor systems

    SciTech Connect

    Borja, R. [Institute of Fat and Its Derivatives (C.S.I.C.), Sevilla (Spain); Banks, C.J. [Environmental Technology Centre, Manchester (United Kingdom)

    1995-03-01

    The anaerobic digestion of ice-cream wastewater, a complex substrate which includes milk proteins, carbohydrates, and lipids, has received little attention. Work using an aerobic contact system showed that at a 7.5-d hydraulic retention time (HRT), with an organic loading rate of 1.7 g COD/Ld and influent TSS (total suspended solids) of 5870 mg/L, the effluent COD was 628 mg/L, BOD was 91 mg/L and TSS was 674. Anaerobic filters have also been used at organic loadings of 6 kg COD/m{sup 3}d applied at a HRT of 0.42 day, with COD removals of 80%. Goodwing showed that this waste was capable of being treated by the UASB process with granulation commencing after 60-70 days, and gas production ranging between 0.73 and 0.93 L CH{sub 4}/g COD removed with loading rates between 0.7 and 3.0 g TOC/Ld. Two-phase anaerobic digestion is an innovative fermentation mode that has recently received increased attention. The kinetically dissimilar fermentation phases, hydrolysis-acidification and acetogenesis-methanation are operated in two separate reactors; the first of which is maintained at a very short HRT. The effluent from the first, acid-forming, phase is used as the substrate for the methane-phase reactor which has a longer HRT or cell immobilization. The aim of this work was to compare the methane production capability and performance of a single-phase upflow fixed bed reactor with a two-phase digestion system. The two-phase digestion system consists of a completely mixed reactor for the acidogenic reaction and an upflow fixed bed reactor for the methanogenic reaction. Because of the high lipid content and COD of ice cream wastewater off site disposal has proved to be both expensive and poses problems to the receiving effluent treatment plant. For this reason the potential for a rapid anaerobic stabilization of the waste, with energy recovery in the form of methane gas, has been investigated in an attempt to minimize plant size and maximize gas production. 9 refs., 2 tabs.

  16. Colliding Beam Fusion Reactor Space Propulsion System

    Microsoft Academic Search

    A. Cheung; M. Binderbauer; F. Liu; A. Qerushi; N. Rostoker; F. J. Wessel

    2004-01-01

    The Colliding Beam Fusion Reactor Space Propulsion System, CBFR-SPS, is an aneutronic, magnetic-field-reversed configuration, fueled by an energetic-ion mixture of hydrogen and boron11 (H-B11). Particle confinement and transport in the CBFR-SPS are classical, hence the system is scaleable. Fusion products are helium ions, alpha-particles, expelled axially out of the system. alpha-particles flowing in one direction are decelerated and their energy

  17. Dynamic Impregnator Reactor System (Poster)

    SciTech Connect

    Not Available

    2012-09-01

    IBRF poster developed for the IBRF showcase. Describes the multifarious system designed for complex feedstock impregnation and processing. IBRF feedstock system has several unit operations combined into one robust system that provides for flexible and staged process configurations, such as spraying, soaking, low-severity pretreatment, enzymatic hydrolysis, fermentation, concentration/evaporation, and distillation.

  18. Hybrid Molten Salt Reactor (HMSR) System Study

    SciTech Connect

    Woolley, Robert D [PPPL; Miller, Laurence F [PPPL

    2014-04-01

    Can the hybrid system combination of (1) a critical fission Molten Salt Reactor (MSR) having a thermal spectrum and a high Conversion Ratio (CR) with (2) an external source of high energy neutrons provide an attractive solution to the world's expanding demand for energy? The present study indicates the answer is an emphatic yes.

  19. The Liquid Annular Reactor System (LARS) propulsion

    Microsoft Academic Search

    J. Powell; H. Ludewig; F. Horn; R. Lenard

    1990-01-01

    A new concept for very high specific impulse (greater than 2,000 seconds) direct nuclear propulsion is described. The concept, termed LARS (Liquid Annular Reactor System) uses liquid nuclear fuel elements to heat hydrogen propellant to very high temperatures (approximately 6,000 K). Operating pressure is moderate (approximately 10 atm), with the result that the outlet hydrogen is virtually 100 percent dissociated

  20. The Liquid Annular Reactor System (LARS) propulsion

    Microsoft Academic Search

    George Maise; Otto W. Lazareth; Frederick Horn; James R. Powell; Hans Ludewig; Roger X. Lenard

    1991-01-01

    A new concept for very high specific impulse (greater than 2000 seconds) direct nuclear propulsion is described. The concept, termed LARS (Liquid Annular Reactor System) uses liquid nuclear fuel elements to heat hydrogen propellant to very high temperatures (about 6000 K). Operating pressure is moderate (about 10 atm), with the result that the outlet hydrogen is virtually 100 percent dissociated

  1. The liquid annular reactor system (LARS) propulsion

    Microsoft Academic Search

    George Maise; Otto W. Lazareth; Frederic Horn; James R. Powell; Hans Ludewig; Roger X. Lenard

    1991-01-01

    A new concept for very high specific impulse (>~2000 seconds) direct nuclear propulsion is described. The concept, termed LARS (Liquid Annular Reactor System) uses liquid nuclear fuel elements to heat hydrogen propellant to very high temperatures (~6000 K). Operating pressure is moderate (~10 atm), with the result that the outlet hydrogen is virtually 100% dissociated to monatomic H. The molten

  2. Limit regulation system for pressurized water nuclear reactors

    Microsoft Academic Search

    W. Aleite; H. W. Bock

    1976-01-01

    Described is a limit regulation system for a pressurized water nuclear reactor in combination with a steam generating system connected to a turbine, the nuclear reactor having control rods as well as an operational regulation system and a protective system, which includes reactor power limiting means operatively associated with the control rods for positioning the same and having response values

  3. Trip setpoint analysis for the reactor protection system of an advanced integral reactor

    Microsoft Academic Search

    Soo Hyung Yang; Soo Hyung Kim; Young Jong Chung; Sung Quun Zee

    2007-01-01

    The trip setpoints for the reactor protection system of a 65-MWt advanced integral reactor have been analyzed through sensitivity evaluations by using the Transients and Setpoint Simulation\\/System-integrated Modular Reactor code. In the analysis, an inadvertent control rod withdrawal event has been considered as an initiating event because this event results in the worst consequences from the viewpoint of the minimum

  4. Gaseous fuel reactor systems for aerospace applications

    NASA Technical Reports Server (NTRS)

    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.

  5. Control system studies for thermionic reactors

    NASA Technical Reports Server (NTRS)

    Hermsen, R. J.; Gronroos, H. G.

    1978-01-01

    In core thermionic reactor concepts are of interest for space missions that require electrical power in the range of a few tens of kilowatts up to several megawatts. The physical principle involved--thermionic direct conversion of heat to electricity at net efficiencies up to 15 percent--offers potential advantages when compared to other nuclear powerplant concepts. However, the integration of the thermionic diode electrode structure with high-temperature nuclear fuel materials presents new design problems and new reactor physical constraints. Among the topics that must be investigated are those associated with the control system. The results of analytical and simulation studies of thermionic reactor control performed at the Jet Propulsion Laboratory are discussed.

  6. Reactor control rod timing system. [LMFBR

    DOEpatents

    Wu, P.T.K.

    1980-03-18

    A fluid driven jet-edge whistle timing system is described for control rods of a nuclear reactor for producing real-time detection of the timing of each control rod in its scram operation. An important parameter in reactor safety, particularly for liquid metal fast breeder reactors (LMFBR), is the time deviation between the time the control rod is released and the time the rod actually reaches the down position. The whistle has a nearly pure tone signal with center frequency (above 100 kHz) far above the frequency band in which the energy of the background noise is concentrated. Each control rod can be fitted with a whistle with a different frequency so that there is no ambiguity in differentiating the signal from each control rod.

  7. Fault-tolerant reactor protection system

    DOEpatents

    Gaubatz, D.C.

    1997-04-15

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

  8. Fault-tolerant reactor protection system

    DOEpatents

    Gaubatz, Donald C. (Cupertino, CA)

    1997-01-01

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

  9. Containment system for supercritical water oxidation reactor

    DOEpatents

    Chastagner, Philippe (3134 Natalie Cir., Augusta, GA 30909-2748)

    1994-01-01

    A system for containment of a supercritical water oxidation reactor in the event of a rupture of the reactor. The system includes a containment for housing the reaction vessel and a communicating chamber for holding a volume of coolant, such as water. The coolant is recirculated and sprayed to entrain and cool any reactants that might have escaped from the reaction vessel. Baffles at the entrance to the chamber prevent the sprayed coolant from contacting the reaction vessel. An impact-absorbing layer is positioned between the vessel and the containment to at least partially absorb momentum of any fragments propelled by the rupturing vessel. Remote, quick-disconnecting fittings exterior to the containment, in cooperation with shut-off valves, enable the vessel to be isolated and the system safely taken off-line. Normally-closed orifices throughout the containment and chamber enable decontamination of interior surfaces when necessary.

  10. Containment system for supercritical water oxidation reactor

    DOEpatents

    Chastagner, P.

    1994-07-05

    A system is described for containment of a supercritical water oxidation reactor in the event of a rupture of the reactor. The system includes a containment for housing the reaction vessel and a communicating chamber for holding a volume of coolant, such as water. The coolant is recirculated and sprayed to entrain and cool any reactants that might have escaped from the reaction vessel. Baffles at the entrance to the chamber prevent the sprayed coolant from contacting the reaction vessel. An impact-absorbing layer is positioned between the vessel and the containment to at least partially absorb momentum of any fragments propelled by the rupturing vessel. Remote, quick-disconnecting fittings exterior to the containment, in cooperation with shut-off valves, enable the vessel to be isolated and the system safely taken off-line. Normally-closed orifices throughout the containment and chamber enable decontamination of interior surfaces when necessary. 2 figures.

  11. The unique safety challenges of space reactor systems

    Microsoft Academic Search

    S. J. Lanes; A. C. Marshall

    1991-01-01

    Compact reactor systems can provide high levels of power for extended periods in space environments. Their relatively low mass and their ability to operate independently of their proximity to the sun make reactor power systems high desirable for many civilian and military space missions. The US Department of Energy is developing reactor system technologies to provide electrical power for space

  12. Integral reactor system and method for fuel cells

    DOEpatents

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

    2013-11-19

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

  13. Transient thermal analysis of a space reactor power system

    Microsoft Academic Search

    M. J. Gaeta; F. R. Best

    1993-01-01

    Space nuclear power systems utilize materials and processes that are completely different from terrestrial reactor systems. Therefore, the tools used to analyze ground-based systems are inappropriate for space reactor design and analysis. The purpose of this study was to develop a space reactor transient analysis tool and to apply this tool to scenarios of interest. The scope of the simulation

  14. Static conversion systems. [for space power reactors

    NASA Technical Reports Server (NTRS)

    Ewell, R.; Mondt, J.

    1985-01-01

    Historically, all space power systems that have actually flown in space have relied on static energy conversion technology. Thus, static conversion is being considered for space nuclear power systems as well. There are four potential static conversion technologies which should be considered. These include: the alkali metal thermoelectric converter (AMTEC), the thermionic converter, the thermoelectric converter, and the thermophotovoltaic converter (TPV). These four conversion technologies will be described in brief detail along with their current status and development needs. In addition, the systems implications of using each of these conversion technologies with a space nuclear reactor power system will be evaluated and some comparisons made.

  15. Plasma generators, reactor systems and related methods

    DOEpatents

    Kong, Peter C. (Idaho Falls, ID); Pink, Robert J. (Pocatello, ID); Lee, James E. (Idaho Falls, ID)

    2007-06-19

    A plasma generator, reactor and associated systems and methods are provided in accordance with the present invention. A plasma reactor may include multiple sections or modules which are removably coupled together to form a chamber. Associated with each section is an electrode set including three electrodes with each electrode being coupled to a single phase of a three-phase alternating current (AC) power supply. The electrodes are disposed about a longitudinal centerline of the chamber and are arranged to provide and extended arc and generate an extended body of plasma. The electrodes are displaceable relative to the longitudinal centerline of the chamber. A control system may be utilized so as to automatically displace the electrodes and define an electrode gap responsive to measure voltage or current levels of the associated power supply.

  16. Staged membrane oxidation reactor system

    DOEpatents

    Repasky, John Michael; Carolan, Michael Francis; Stein, VanEric Edward; Chen, Christopher Ming-Poh

    2013-04-16

    Ion transport membrane oxidation system comprising (a) two or more membrane oxidation stages, each stage comprising a reactant zone, an oxidant zone, one or more ion transport membranes separating the reactant zone from the oxidant zone, a reactant gas inlet region, a reactant gas outlet region, an oxidant gas inlet region, and an oxidant gas outlet region; (b) an interstage reactant gas flow path disposed between each pair of membrane oxidation stages and adapted to place the reactant gas outlet region of a first stage of the pair in flow communication with the reactant gas inlet region of a second stage of the pair; and (c) one or more reactant interstage feed gas lines, each line being in flow communication with any interstage reactant gas flow path or with the reactant zone of any membrane oxidation stage receiving interstage reactant gas.

  17. Staged membrane oxidation reactor system

    DOEpatents

    Repasky, John Michael; Carolan, Michael Francis; Stein, VanEric Edward; Chen, Christopher Ming-Poh

    2014-05-20

    Ion transport membrane oxidation system comprising (a) two or more membrane oxidation stages, each stage comprising a reactant zone, an oxidant zone, one or more ion transport membranes separating the reactant zone from the oxidant zone, a reactant gas inlet region, a reactant gas outlet region, an oxidant gas inlet region, and an oxidant gas outlet region; (b) an interstage reactant gas flow path disposed between each pair of membrane oxidation stages and adapted to place the reactant gas outlet region of a first stage of the pair in flow communication with the reactant gas inlet region of a second stage of the pair; and (c) one or more reactant interstage feed gas lines, each line being in flow communication with any interstage reactant gas flow path or with the reactant zone of any membrane oxidation stage receiving interstage reactant gas.

  18. Staged membrane oxidation reactor system

    DOEpatents

    Repasky, John Michael; Carolan, Michael Francis; Stein, VanEric Edward; Chen, Christopher Ming-Poh

    2012-09-11

    Ion transport membrane oxidation system comprising (a) two or more membrane oxidation stages, each stage comprising a reactant zone, an oxidant zone, one or more ion transport membranes separating the reactant zone from the oxidant zone, a reactant gas inlet region, a reactant gas outlet region, an oxidant gas inlet region, and an oxidant gas outlet region; (b) an interstage reactant gas flow path disposed between each pair of membrane oxidation stages and adapted to place the reactant gas outlet region of a first stage of the pair in flow communication with the reactant gas inlet region of a second stage of the pair; and (c) one or more reactant interstage feed gas lines, each line being in flow communication with any interstage reactant gas flow path or with the reactant zone of any membrane oxidation stage receiving interstage reactant gas.

  19. Scaling laws for modeling nuclear reactor systems

    Microsoft Academic Search

    A. N. Nahavandi; F. S. Castellana; E. N. Moradkhanian

    1979-01-01

    Scale models are used to predict the behavior of nuclear reactor systems during normal and abnormal operation as well as under accident conditions. Three types of scaling procedures are considered: time-reducing, time-preserving volumetric, and time-preserving idealized model\\/prototype. The necessary relations between the model and the full-scale unit are developed for each scaling type. Based on these relationships, it is shown

  20. Colliding beam fusion reactor space propulsion system

    Microsoft Academic Search

    Frank J. Wessel; Michl W. Binderbauer; Norman Rostoker; Hafiz Ur Rahman; Joseph O'Toole

    2000-01-01

    We describe a space propulsion system based on the Colliding Beam Fusion Reactor (CBFR). The CBFR is a high-beta, field-reversed, magnetic configuration with ion energies in the range of hundreds of keV. Repetitively-pulsed ion beams sustain the plasma distribution and provide current drive. The confinement physics is based on the Vlasov-Maxwell equation, including a Fokker Planck collision operator and all

  1. Power conditioning for space nuclear reactor systems

    NASA Technical Reports Server (NTRS)

    Berman, Baruch

    1987-01-01

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

  2. Reactor protection system design alternatives for sodium fast reactors

    E-print Network

    DeWitte, Jacob D. (Jacob Dominic)

    2011-01-01

    Historically, unprotected transients have been viewed as design basis events that can significantly challenge sodium-cooled fast reactors. The perceived potential consequences of a severe unprotected transient in a ...

  3. Standard Operating Procedure (Polypropylene Reactor System)

    E-print Network

    Choi, Kyu Yong

    rotation speed). 2. Prepare the high pressure glass reactor inside of the glove box: a. Fill the reactor, collect the reaction mixture in a glass beaker and add acidic methanol. 8. Clean out and dry the reactor

  4. Integrated application of upflow anaerobic sludge blanket reactor for the treatment of wastewaters.

    PubMed

    Latif, Muhammad Asif; Ghufran, Rumana; Wahid, Zularisam Abdul; Ahmad, Anwar

    2011-10-15

    The UASB process among other treatment methods has been recognized as a core method of an advanced technology for environmental protection. This paper highlights the treatment of seven types of wastewaters i.e. palm oil mill effluent (POME), distillery wastewater, slaughterhouse wastewater, piggery wastewater, dairy wastewater, fishery wastewater and municipal wastewater (black and gray) by UASB process. The purpose of this study is to explore the pollution load of these wastewaters and their treatment potential use in upflow anaerobic sludge blanket process. The general characterization of wastewater, treatment in UASB reactor with operational parameters and reactor performance in terms of COD removal and biogas production are thoroughly discussed in the paper. The concrete data illustrates the reactor configuration, thus giving maximum awareness about upflow anaerobic sludge blanket reactor for further research. The future aspects for research needs are also outlined. PMID:21764417

  5. The Liquid Annular Reactor System (LARS) propulsion

    NASA Astrophysics Data System (ADS)

    Maise, George; Lazareth, Otto W.; Horn, Frederick; Powell, James R.; Ludewig, Hans; Lenard, Roger X.

    A new concept for very high specific impulse (greater than 2000 seconds) direct nuclear propulsion is described. The concept, termed LARS (Liquid Annular Reactor System) uses liquid nuclear fuel elements to heat hydrogen propellant to very high temperatures (about 6000 K). Operating pressure is moderate (about 10 atm), with the result that the outlet hydrogen is virtually 100 percent dissociated to monatomic H. The molten fuel is contained in a solid container of its own material, which is rotated to stabilize the liquid layer by centripetal force. LARS reactor designs are described, together with neutronic and thermal-hydraulic analyses. Power levels are on the order of 200 megawatts. Typically, LARS designs use 7 rotating fuel elements, are beryllium moderated and have critical radii of about 100 cm.

  6. The liquid annular reactor system (LARS) propulsion

    SciTech Connect

    Maise, G.; Lazareth, O.W.; Horn, F.; Powell, J.R.; Ludewig, H. (Brookhaven National Laboratory, Department of Nuclear Energy, Upton, NY (USA)); Lenard, R.X. (SDIO The Pentagon, Washington, DC (USA))

    1991-01-05

    A new concept for very high specific impulse ({gt}2000 seconds) direct nuclear propulsion is described. The concept, termed LARS (Liquid Annular Reactor System) uses liquid nuclear fuel elements to heat hydrogen propellant to very high temperatures ({similar to}6000 K). Operating pressure is moderate ({similar to}10 atm), with the result that the outlet hydrogen is virtually 100% dissociated to monatomic H. The molten fuel is contained in a solid container of its own material, which is rotated to stabilize the liquid layer by centripetal force. LARS reactor designs are described, together with neutronic and thermal-hydraulic analyses. Power levels are on the order of 200 megawatts. Typically, LARS designs use 7 rotating fuel elements, are beryllium moderated and have critical radii of {similar to}100 cm (core L/D{approx}1.5).

  7. The Liquid Annular Reactor System (LARS) propulsion

    NASA Technical Reports Server (NTRS)

    Powell, James; Ludewig, Hans; Horn, Frederick; Lenard, Roger

    1990-01-01

    A concept for very high specific impulse (greater than 2000 seconds) direct nuclear propulsion is described. The concept, termed the liquid annular reactor system (LARS), uses liquid nuclear fuel elements to heat hydrogen propellant to very high temperatures (approximately 6000 K). Operating pressure is moderate (approximately 10 atm), with the result that the outlet hydrogen is virtually 100 percent dissociated to monatomic H. The molten fuel is contained in a solid container of its own material, which is rotated to stabilize the liquid layer by centripetal force. LARS reactor designs are described, together with neutronic and thermal-hydraulic analyses. Power levels are on the order of 200 megawatts. Typically, LARS designs use seven rotating fuel elements, are beryllium moderated, and have critical radii of approximately 100 cm (core L/D approximately equal to 1.5).

  8. The Liquid Annular Reactor System (LARS) propulsion

    NASA Astrophysics Data System (ADS)

    Powell, J.; Ludewig, H.; Horn, F.; Lenard, R.

    A new concept for very high specific impulse (greater than 2,000 seconds) direct nuclear propulsion is described. The concept, termed LARS (Liquid Annular Reactor System) uses liquid nuclear fuel elements to heat hydrogen propellant to very high temperatures (approximately 6,000 K). Operating pressure is moderate (approximately 10 atm), with the result that the outlet hydrogen is virtually 100 percent dissociated to monatomic H. The molten fuel is contained in a solid container of its own material, which is rotated to stabilize the liquid layer by centripetal force. LARS reactor designs are described, together with neutronic and thermal-hydraulic analyses. Power levels are on the order of 200 megawatts. Typically, LARS designs use 7 rotating fuel elements, are beryllium moderated and have critical radii of approximately 100 cm (core L/D is approximately 1.5).

  9. The liquid annular reactor system (LARS) propulsion

    NASA Astrophysics Data System (ADS)

    Maise, George; Lazareth, Otto W.; Horn, Frederic; Powell, James R.; Ludewig, Hans; Lenard, Roger X.

    1991-01-01

    A new concept for very high specific impulse (?2000 seconds) direct nuclear propulsion is described. The concept, termed LARS (Liquid Annular Reactor System) uses liquid nuclear fuel elements to heat hydrogen propellant to very high temperatures (˜6000 K). Operating pressure is moderate (˜10 atm), with the result that the outlet hydrogen is virtually 100% dissociated to monatomic H. The molten fuel is contained in a solid container of its own material, which is rotated to stabilize the liquid layer by centripetal force. LARS reactor designs are described, together with neutronic and thermal-hydraulic analyses. Power levels are on the order of 200 megawatts. Typically, LARS designs use 7 rotating fuel elements, are beryllium moderated and have critical radii of ˜100 cm (core L/D?1.5).

  10. Fuel handling system for a nuclear reactor

    DOEpatents

    Saiveau, James G. (Hickory Hills, IL); Kann, William J. (Park Ridge, IL); Burelbach, James P. (Glen Ellyn, IL)

    1986-01-01

    A pool type nuclear fission reactor has a core, with a plurality of core elements and a redan which confines coolant as a hot pool at a first end of the core separated from a cold pool at a second end of the core by the redan. A fuel handling system for use with such reactors comprises a core element storage basket located outside of the redan in the cold pool. An access passage is formed in the redan with a gate for opening and closing the passage to maintain the temperature differential between the hot pool and the cold pool. A mechanism is provided for opening and closing the gate. A lifting arm is also provided for manipulating the fuel core elements through the access passage between the storage basket and the core when the redan gate is open.

  11. Passive compact molten salt reactor (PCMSR), modular thermal breeder reactor with totally passive safety system

    NASA Astrophysics Data System (ADS)

    Harto, Andang Widi

    2012-06-01

    Design Study Passive Compact Molten Salt Reactor (PCMSR) with totally passive safety system has been performed. The term of Compact in the PCMSR name means that the reactor system is designed to have relatively small volume per unit power output by using modular and integral concept. In term of modular, the reactor system consists of three modules, i.e. reactor module, turbine module and fuel management module. The reactor module is an integral design that consists of reactor, primary and intermediate heat exchangers and passive post shutdown cooling system. The turbine module is an integral design of a multi heating, multi cooling, regenerative gas turbine. The fuel management module consists of all equipments related to fuel preparation, fuel reprocessing and radioactive handling. The preliminary calculations show that the PCMSR has negative temperature and void reactivity coefficient, passive shutdown characteristic related to fuel pump failure and possibility of using natural circulation for post shutdown cooling system.

  12. Passive compact molten salt reactor (PCMSR), modular thermal breeder reactor with totally passive safety system

    SciTech Connect

    Harto, Andang Widi [Engineering Physics Department, Faculty of Engineering, Gadjah Mada University (Indonesia)

    2012-06-06

    Design Study Passive Compact Molten Salt Reactor (PCMSR) with totally passive safety system has been performed. The term of Compact in the PCMSR name means that the reactor system is designed to have relatively small volume per unit power output by using modular and integral concept. In term of modular, the reactor system consists of three modules, i.e. reactor module, turbine module and fuel management module. The reactor module is an integral design that consists of reactor, primary and intermediate heat exchangers and passive post shutdown cooling system. The turbine module is an integral design of a multi heating, multi cooling, regenerative gas turbine. The fuel management module consists of all equipments related to fuel preparation, fuel reprocessing and radioactive handling. The preliminary calculations show that the PCMSR has negative temperature and void reactivity coefficient, passive shutdown characteristic related to fuel pump failure and possibility of using natural circulation for post shutdown cooling system.

  13. Systems analysis of the CANDU 3 Reactor

    SciTech Connect

    Wolfgong, J.R.; Linn, M.A.; Wright, A.L.; Olszewski, M.; Fontana, M.H. [Oak Ridge National Lab., TN (United States)

    1993-07-01

    This report presents the results of a systems failure analysis study of the CANDU 3 reactor design; the study was performed for the US Nuclear Regulatory Commission. As part of the study a review of the CANDU 3 design documentation was performed, a plant assessment methodology was developed, representative plant initiating events were identified for detailed analysis, and a plant assessment was performed. The results of the plant assessment included classification of the CANDU 3 event sequences that were analyzed, determination of CANDU 3 systems that are ``significant to safety,`` and identification of key operator actions for the analyzed events.

  14. Development of a system model for advanced small modular reactors.

    SciTech Connect

    Lewis, Tom Goslee,; Holschuh, Thomas Vernon,

    2014-01-01

    This report describes a system model that can be used to analyze three advance small modular reactor (SMR) designs through their lifetime. Neutronics of these reactor designs were evaluated using Monte Carlo N-Particle eXtended (MCNPX/6). The system models were developed in Matlab and Simulink. A major thrust of this research was the initial scoping analysis of Sandia's concept of a long-life fast reactor (LLFR). The inherent characteristic of this conceptual design is to minimize the change in reactivity over the lifetime of the reactor. This allows the reactor to operate substantially longer at full power than traditional light water reactors (LWRs) or other SMR designs (e.g. high temperature gas reactor (HTGR)). The system model has subroutines for lifetime reactor feedback and operation calculations, thermal hydraulic effects, load demand changes and a simplified SCO2 Brayton cycle for power conversion.

  15. Granules characteristics in the vertical profile of a full-scale upflow anaerobic sludge blanket reactor treating poultry slaughterhouse wastewater

    Microsoft Academic Search

    Valéria Del Nery; Eloisa Pozzi; Márcia H. R. Z. Damianovic; Mércia R. Domingues; Marcelo Zaiat

    2008-01-01

    The performance and the granules characteristics of a 450m3 -UASB reactor operating for 1228days, treating poultry slaughterhouse wastewater with an average COD reduction of 85% was examined. Granules were sampled in three different positions along the vertical central line of the reactor, revealing variations in the concentration of volatile total solids. Although the reactor had been in operation for an

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

    DOEpatents

    Hunsbedt, Anstein (Los Gatos, CA); Boardman, Charles E. (Saratoga, CA); Hui, Marvin M. (Sunnyvale, CA); Berglund, Robert C. (Saratoga, CA)

    1991-01-01

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

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

    DOEpatents

    Hunsbedt, Anstein (Los Gatos, CA); Boardman, Charles E. (Saratoga, CA)

    1990-01-01

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

  18. Standard Operating Procedure (Microchannel Reactor System)

    E-print Network

    Choi, Kyu Yong

    . 6. Connect the glass reactor to the HPLC pump. 7. Connect the glass reservoir to the outlet to the reaction pressure. 8. Turn on the HPLC pump to start the flow of solvent into the microchannel reactor. 9 toluene to the glass reactor and connect to the HPLC pump. b. Turn on the HPLC pump to purge the reactor

  19. Evaluation of performance in a combined UASB and aerobic contact oxidation process treating acrylic wastewater.

    PubMed

    Li, Anfeng; Dong, Na; He, Manni; Pan, Tao

    2015-04-01

    The lab-scale and full-scale performance of a combined mesophilic up-flow anaerobic sludge blanket (UASB) and aerobic contact oxidation (ACO) process for treating acrylic wastewater was studied. During lab-scale experiment, the overwhelmed volumetric load for UASB was above 6?kg chemical oxygen demand (COD) ·(m(-3)·d(-1)) since COD removal efficiency dropped dramatically from 73% at 6?kg COD·(m(-3)·d(-1)) to 61% at 7?kg COD·(m(-3)·d(-1)) and 53% at 8?kg COD·(m(-3)·d(-1)). Further results showed that an up-flow fluid velocity of 0.5?m?h(-1) for UASB obtained a highest COD removal efficiency of 75%, and the optimum COD volumetric load for the corresponding ACO was 1.00?kg COD·(m(-3)·d(-1)). Based on the configuration of the lab-scale experiment, a full-scale application with an acrylic wastewater treatment capacity of 8?m(3)?h(-1) was constructed and operated at a volumetric load of 5.5?kg COD·(m(-3)·d(-1)), an up-flow fluid velocity of 0.5?m?h(-1) for UASB and a volumetric load of 0.9?kg COD·(m(-3)·d(-1)) for ACO; and the final effluent COD was around 740?mg?L(-1). The results suggest that a combined UASB-ACO process is promising for treating acrylic wastewater. PMID:25204720

  20. Integrated systems analysis of the PIUS reactor

    SciTech Connect

    Fullwood, F.; Kroeger, P.; Higgins, J. [Brookhaven National Lab., Upton, NY (United States)] [and others

    1993-11-01

    Results are presented of a systems failure analysis of the PIUS plant systems that are used during normal reactor operation and postulated accidents. This study was performed to provide the NRC with an understanding of the behavior of the plant. The study applied two diverse failure identification methods, Failure Modes Effects & Criticality Analysis (FMECA) and Hazards & Operability (HAZOP) to the plant systems, supported by several deterministic analyses. Conventional PRA methods were also used along with a scheme for classifying events by initiator frequency and combinations of failures. Principal results of this study are: (a) an extensive listing of potential event sequences, grouped in categories that can be used by the NRC, (b) identification of support systems that are important to safety, and (c) identification of key operator actions.

  1. AN INTEGRATED SYSTEMS CFD SIMULATION OF A PEBBLE BED REACTOR

    Microsoft Academic Search

    PG Rousseau; WA Landman

    The theoretical basis of a systems CFD model of a pebble bed reactor is discussed. This model is employed to simulate the thermal-fluid phenomena of the reactor core. The formulation of the fundamental equations results in a collection of one-dimensional elements that can be used to construct a network model of the reactor. One preliminary test is discussed to illustrate

  2. Nuclear reactor fuel rod attachment system

    DOEpatents

    Not Available

    1980-09-17

    A reusable system is described for removably attaching a nuclear reactor fuel rod to a support member. A locking cap is secured to the fuel rod and a locking strip is fastened to the support member. The locking cap has two opposing fingers shaped to form a socket having a body portion. The locking strip has an extension shaped to rigidly attach to the socket's body portion. The locking cap's fingers are resiliently deflectable. For attachment, the locking cap is longitudinally pushed onto the locking strip causing the extension to temporarily deflect open the fingers to engage the socket's body portion. For removal, the process is reversed.

  3. Structural materials challenges for advanced reactor systems

    NASA Astrophysics Data System (ADS)

    Yvon, P.; Carré, F.

    2009-03-01

    Key technologies for advanced nuclear systems encompass high temperature structural materials, fast neutron resistant core materials, and specific reactor and power conversion technologies (intermediate heat exchanger, turbo-machinery, high temperature electrolytic or thermo-chemical water splitting processes, etc.). The main requirements for the materials to be used in these reactor systems are dimensional stability under irradiation, whether under stress (irradiation creep or relaxation) or without stress (swelling, growth), an acceptable evolution under ageing of the mechanical properties (tensile strength, ductility, creep resistance, fracture toughness, resilience) and a good behavior in corrosive environments (reactor coolant or process fluid). Other criteria for the materials are their cost to fabricate and to assemble, and their composition could be optimized in order for instance to present low-activation (or rapid desactivation) features which facilitate maintenance and disposal. These requirements have to be met under normal operating conditions, as well as in incidental and accidental conditions. These challenging requirements imply that in most cases, the use of conventional nuclear materials is excluded, even after optimization and a new range of materials has to be developed and qualified for nuclear use. This paper gives a brief overview of various materials that are essential to establish advanced systems feasibility and performance for in pile and out of pile applications, such as ferritic/martensitic steels (9-12% Cr), nickel based alloys (Haynes 230, Inconel 617, etc.), oxide dispersion strengthened ferritic/martensitic steels, and ceramics (SiC, TiC, etc.). This article gives also an insight into the various natures of R&D needed on advanced materials, including fundamental research to investigate basic physical and chemical phenomena occurring in normal and accidental operating conditions, lab-scale tests to characterize candidate materials mechanical properties and corrosion resistance, as well as component mock-up tests on technology loops to validate potential applications while accounting for mechanical design rules and manufacturing processes. The selection, assessment and validation of materials necessitate a large number of experiments, involving rare and expensive facilities such as research reactors, hot laboratories or corrosion loops. The modelling and the codification of the behaviour of materials will always involve the use of such technological experiments, but it is of utmost importance to develop also a predictive material science. Finally, the paper stresses the benefit of prospects of multilateral collaboration to join skills and share efforts of R&D to achieve in the nuclear field breakthroughs on materials that have already been achieved over the past decades in other industry sectors (aeronautics, metallurgy, chemistry, etc.).

  4. Systems aspects of a space nuclear reactor power system

    NASA Technical Reports Server (NTRS)

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

    1988-01-01

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

  5. Nuclear reactor cooling system decontamination reagent regeneration

    DOEpatents

    Anstine, Larry D. (San Jose, CA); James, Dean B. (Saratoga, CA); Melaika, Edward A. (Berkeley, CA); Peterson, Jr., John P. (Livermore, CA)

    1985-01-01

    An improved method for decontaminating the coolant system of water-cooled nuclear power reactors and for regenerating the decontamination solution. A small amount of one or more weak-acid organic complexing agents is added to the reactor coolant, and the pH is adjusted to form a decontamination solution which is circulated throughout the coolant system to dissolve metal oxides from the interior surfaces and complex the resulting metal ions and radionuclide ions. The coolant containing the complexed metal ions and radionuclide ions is passed through a strong-base anion exchange resin bed which has been presaturated with a solution containing the complexing agents in the same ratio and having the same pH as the decontamination solution. As the decontamination solution passes through the resin bed, metal-complexed anions are exchanged for the metal-ion-free anions on the bed, while metal-ion-free anions in the solution pass through the bed, thus removing the metal ions and regenerating the decontamination solution.

  6. Proceedings of a Symposium on Advanced Compact Reactor Systems

    NASA Astrophysics Data System (ADS)

    1983-04-01

    Reactor system technologies suitable for a variety of aerospace and terrestrial applications are considered. Technologies, safety and regulatory considerations, potential applications, and research and development opportunities are covered.

  7. Proceedings of a Symposium on Advanced Compact Reactor Systems

    NASA Technical Reports Server (NTRS)

    1983-01-01

    Reactor system technologies suitable for a variety of aerospace and terrestrial applications are considered. Technologies, safety and regulatory considerations, potential applications, and research and development opportunities are covered.

  8. Auxiliary reactor for a hydrocarbon reforming system

    DOEpatents

    Clawson, Lawrence G.; Dorson, Matthew H.; Mitchell, William L.; Nowicki, Brian J.; Bentley, Jeffrey M.; Davis, Robert; Rumsey, Jennifer W.

    2006-01-17

    An auxiliary reactor for use with a reformer reactor having at least one reaction zone, and including a burner for burning fuel and creating a heated auxiliary reactor gas stream, and heat exchanger for transferring heat from auxiliary reactor gas stream and heat transfer medium, preferably two-phase water, to reformer reaction zone. Auxiliary reactor may include first cylindrical wall defining a chamber for burning fuel and creating a heated auxiliary reactor gas stream, the chamber having an inlet end, an outlet end, a second cylindrical wall surrounding first wall and a second annular chamber there between. The reactor being configured so heated auxiliary reactor gas flows out the outlet end and into and through second annular chamber and conduit which is disposed in second annular chamber, the conduit adapted to carry heat transfer medium and being connectable to reformer reaction zone for additional heat exchange.

  9. Application of Stable Adaptive Schemes to Nuclear Reactor Systems, (II)

    Microsoft Academic Search

    Toshio FUKUDA

    1979-01-01

    The parameter identification and adaptive control schemes applied in a previous study to a nonlinear point reactor are extended to the case of a loosely-coupled-core reactor with internal feedbacks, constituting a nonlinear overall system. Both schemes are shown to be stable, with the system newly represented on the pattern of the Model Reference Adaptive System (MRAS) with use made of

  10. Sliding mode control of the space nuclear reactor system

    Microsoft Academic Search

    Y. B. Shtessel

    1998-01-01

    The automatic control system (ACS) of the space nuclear reactor power system TOPAZ II that generates electricity from nuclear heat using in-core thermionic converters is considered. Sliding mode control technique is applied to the reactor system controller design in order to improve robustness and accuracy of tracking of a thermal power reference profile in a start-up regime and a payload

  11. REACTOR - a Concept for establishing a System-of-Systems

    NASA Astrophysics Data System (ADS)

    Haener, Rainer; Hammitzsch, Martin; Wächter, Joachim

    2014-05-01

    REACTOR is a working title for activities implementing reliable, emergent, adaptive, and concurrent collaboration on the basis of transactional object repositories. It aims at establishing federations of autonomous yet interoperable systems (Systems-of-Systems), which are able to expose emergent behaviour. Following the principles of event-driven service-oriented architectures (SOA 2.0), REACTOR enables adaptive re-organisation by dynamic delegation of responsibilities and novel yet coherent monitoring strategies by combining information from different domains. Thus it allows collaborative decision-processes across system, discipline, and administrative boundaries. Interoperability is based on two approaches that implement interconnection and communication between existing heterogeneous infrastructures and information systems: Coordinated (orchestration-based) communication and publish/subscribe (choreography-based) communication. Choreography-based communication ensures the autonomy of the participating systems to the highest possible degree but requires the implementation of adapters, which provide functional access to information (publishing/consuming events) via a Message Oriented Middleware (MOM). Any interconnection of the systems (composition of service and message cascades) is established on the basis of global conversations that are enacted by choreographies specifying the expected behaviour of the participating systems with respect to agreed Service Level Agreements (SLA) required by e.g. national authorities. The specification of conversations, maintained in commonly available repositories also enables the utilisation of systems for purposes (evolving) other than initially intended. Orchestration-based communication additionally requires a central component that controls the information transfer via service requests or event processing and also takes responsibility of managing business processes. Commonly available transactional object repositories are well suited to establish brokers, which mediate metadata and semantic information about the resources of all involved systems. This concept has been developed within the project Collaborative, Complex, and Critical Decision-Support in Evolving Crises (TRIDEC) on the basis of semantic registries describing all facets of events and services utilisable for crisis management systems. The implementation utilises an operative infrastructure including an Enterprise Service Bus (ESB), adapters to proprietary sensor systems, a workflow engine, and a broker-based MOM. It also applies current technologies like actor-based frameworks for highly concurrent, distributed, and fault tolerant event-driven applications. Therefore REACTOR implementations are well suited to be hosted in a cloud that provides Infrastructure as a Service (IaaS). To provide low entry barriers for legacy and future systems, REACTOR adapts the principles of Design by Contract (DbC) as well as standardised and common information models like the Sensor Web Enablement (SWE) or the JavaScript Object Notation for geographic features (GeoJSON). REACTOR has been applied exemplarily within two different scenarios, Natural Crisis Management and Industrial Subsurface Development.

  12. Fluid sampling system for a nuclear reactor

    DOEpatents

    Lau, Louis K. (Monroeville, PA); Alper, Naum I. (Monroeville, PA)

    1994-01-01

    A system of extracting fluid samples, either liquid or gas, from the interior of a nuclear reactor containment utilizes a jet pump. To extract the sample fluid, a nonradioactive motive fluid is forced through the inlet and discharge ports of a jet pump located outside the containment, creating a suction that draws the sample fluid from the containment through a sample conduit connected to the pump suction port. The mixture of motive fluid and sample fluid is discharged through a return conduit to the interior of the containment. The jet pump and means for removing a portion of the sample fluid from the sample conduit can be located in a shielded sample grab station located next to the containment. A non-nuclear grade active pump can be located outside the grab sampling station and the containment to pump the nonradioactive motive fluid through the jet pump.

  13. Fluid sampling system for a nuclear reactor

    DOEpatents

    Lau, L.K.; Alper, N.I.

    1994-11-22

    A system of extracting fluid samples, either liquid or gas, from the interior of a nuclear reactor containment utilizes a jet pump. To extract the sample fluid, a nonradioactive motive fluid is forced through the inlet and discharge ports of a jet pump located outside the containment, creating a suction that draws the sample fluid from the containment through a sample conduit connected to the pump suction port. The mixture of motive fluid and sample fluid is discharged through a return conduit to the interior of the containment. The jet pump and means for removing a portion of the sample fluid from the sample conduit can be located in a shielded sample grab station located next to the containment. A non-nuclear grade active pump can be located outside the grab sampling station and the containment to pump the nonradioactive motive fluid through the jet pump. 1 fig.

  14. Nuclear reactor fuel rod attachment system

    DOEpatents

    Christiansen, David W. (Kennewick, WA)

    1982-01-01

    A reusable system for removably attaching a nuclear reactor fuel rod (12) to a support member (14). A locking cap (22) is secured to the fuel rod (12) and a locking strip (24) is fastened to the support member (14). The locking cap (22) has two opposing fingers (24a and 24b) shaped to form a socket having a body portion (26). The locking strip has an extension (36) shaped to rigidly attach to the socket's body portion (26). The locking cap's fingers are resiliently deflectable. For attachment, the locking cap (22) is longitudinally pushed onto the locking strip (24) causing the extension (36) to temporarily deflect open the fingers (24a and 24b) to engage the socket's body portion (26). For removal, the process is reversed.

  15. Liquid metal cooled nuclear reactors with passive cooling system

    DOEpatents

    Hunsbedt, Anstein (Los Gatos, CA); Fanning, Alan W. (San Jose, CA)

    1991-01-01

    A liquid metal cooled nuclear reactor having a passive cooling system for removing residual heat resulting from fuel decay during reactor shutdown. The passive cooling system comprises a plurality of cooling medium flow circuits which cooperate to remove and carry heat away from the fuel core upon loss of the normal cooling flow circuit to areas external thereto.

  16. Design of a 25-kWe Surface Reactor System Based on SNAP Reactor Technologies

    SciTech Connect

    Dixon, David D. [Nuclear Systems Design Group, Los Alamos National Laboratory, Los Alamos, NM 87545 (United States); Department of Nuclear Engineering, North Carolina State University, Raleigh, NC 27695 (United States); Hiatt, Matthew T. [Nuclear Systems Design Group, Los Alamos National Laboratory, Los Alamos, NM 87545 (United States); Department of Nuclear Engineering, Texas A and M University, College Station, TX 77843 (United States); Poston, David I.; Kapernick, Richard J. [Nuclear Systems Design Group, Los Alamos National Laboratory, Los Alamos, NM 87545 (United States)

    2006-01-20

    A Hastelloy-X clad, sodium-potassium (NaK-78) cooled, moderated spectrum reactor using uranium zirconium hydride (UZrH) fuel based on the SNAP program reactors is a promising design for use in surface power systems. This paper presents a 98 kWth reactor for a power system the uses multiple Stirling engines to produce 25 kWe-net for 5 years. The design utilizes a pin type geometry containing UZrHx fuel clad with Hastelloy-X and NaK-78 flowing around the pins as coolant. A compelling feature of this design is its use of 49.9% enriched U, allowing it to be classified as a category III-D attractiveness and reducing facility costs relative to highly-enriched space reactor concepts. Presented below are both the design and an analysis of this reactor's criticality under various safety and operations scenarios.

  17. Design of a 25-kWe Surface Reactor System Based on SNAP Reactor Technologies

    NASA Astrophysics Data System (ADS)

    Dixon, David D.; Hiatt, Matthew T.; Poston, David I.; Kapernick, Richard J.

    2006-01-01

    A Hastelloy-X clad, sodium-potassium (NaK-78) cooled, moderated spectrum reactor using uranium zirconium hydride (UZrH) fuel based on the SNAP program reactors is a promising design for use in surface power systems. This paper presents a 98 kWth reactor for a power system the uses multiple Stirling engines to produce 25 kWe-net for 5 years. The design utilizes a pin type geometry containing UZrHx fuel clad with Hastelloy-X and NaK-78 flowing around the pins as coolant. A compelling feature of this design is its use of 49.9% enriched U, allowing it to be classified as a category III-D attractiveness and reducing facility costs relative to highly-enriched space reactor concepts. Presented below are both the design and an analysis of this reactor's criticality under various safety and operations scenarios.

  18. 78 FR 64027 - Preoperational Testing of Emergency Core Cooling Systems for Pressurized-Water Reactors

    Federal Register 2010, 2011, 2012, 2013, 2014

    2013-10-25

    ...ECCSs) of pressurized water reactors (PWRs...Advanced Pressurized-Water Reactor, U.S. Evolutionary Power Reactor, and AP1000...Criteria for Nuclear Power Plants and Fuel Reprocessing...Systems for Pressurized-Water Reactors.''...

  19. Exhaust system with emissions storage device and plasma reactor

    DOEpatents

    Hoard, John W. (Livonia, MI)

    1998-01-01

    An exhaust system for a combustion system, comprising a storage device for collecting NO.sub.x, hydrocarbon, or particulate emissions, or mixture of these emissions, and a plasma reactor for destroying the collected emissions is described. After the emission is collected in by the storage device for a period of time, the emission is then destroyed in a non-thermal plasma generated by the plasma reactor. With respect to the direction of flow of the exhaust stream, the storage device must be located before the terminus of the plasma reactor, and it may be located wholly before, overlap with, or be contained within the plasma reactor.

  20. Applications of plasma core reactors to terrestrial energy systems

    NASA Technical Reports Server (NTRS)

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

    1974-01-01

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

  1. Computer optimization of reactor-thermoelectric space power systems

    NASA Technical Reports Server (NTRS)

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

    1973-01-01

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

  2. Thermionic switched self-actuating reactor shutdown system

    DOEpatents

    Barrus, Donald M. (San Jose, CA); Shires, Charles D. (San Jose, CA); Brummond, William A. (Livermore, CA)

    1989-01-01

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

  3. Technology gap analysis on sodium-cooled reactor fuel handling system supporting advanced burner reactor development.

    SciTech Connect

    Chikazawa, Y.; Farmer, M.; Grandy, C.; Nuclear Engineering Division

    2009-03-01

    The goals of the Global Nuclear Energy Partnership (GNEP) are to expand the use of nuclear energy to meet increasing global energy demand in an environmentally sustainable manner, to address nuclear waste management issues without making separated plutonium, and to address nonproliferation concerns. The advanced burner reactor (ABR) is a fast reactor concept which supports the GNEP fuel cycle system. Since the integral fast reactor (IFR) and advanced liquid-metal reactor (ALMR) projects were terminated in 1994, there has been no major development on sodium-cooled fast reactors in the United States. Therefore, in support of the GNEP fast reactor program, the history of sodium-cooled reactor development was reviewed to support the initiation of this technology within the United States and to gain an understanding of the technology gaps that may still remain for sodium fast reactor technology. The fuel-handling system is a key element of any fast reactor design. The major functions of this system are to receive, test, store, and then load fresh fuel into the core; unload from the core; then clean, test, store, and ship spent fuel. Major requirements are that the system must be reliable and relatively easy to maintain. In addition, the system should be designed so that it does not adversely impact plant economics from the viewpoints of capital investment or plant operations. In this gap analysis, information on fuel-handling operating experiences in the following reactor plants was carefully reviewed: EBR-I, SRE, HNPF, Fermi, SEFOR, FFTF, CRBR, EBR-II, DFR, PFR, Rapsodie, Phenix, Superphenix, KNK, SNR-300, Joyo, and Monju. The results of this evaluation indicate that a standardized fuel-handling system for a commercial fast reactor is yet to be established. However, in the past sodium-cooled reactor plants, most major fuel-handling components-such as the rotatable plug, in-vessel fuel-handling machine, ex-vessel fuel transportation cask, ex-vessel sodium-cooled storage, and cleaning stations-have accumulated satisfactory construction and operation experiences. In addition, two special issues for future development are described in this report: large capacity interim storage and transuranic-bearing fuel handling.

  4. Anaerobic biogranulation in a hybrid reactor treating phenolic waste.

    PubMed

    Ramakrishnan, Anushyaa; Gupta, S K

    2006-10-11

    Granulation was examined in four similar anaerobic hybrid reactors 15.5L volume (with an effective volume of 13.5L) during the treatment of synthetic coal wastewater at the mesophilic temperature of 27+/-5 degrees C. The hybrid reactors are a combination of UASB unit at the lower part and an anaerobic filter at the upper end. Synthetic wastewater with an average chemical oxygen demand (COD) of 2,240 mg/L, phenolics concentration of 752 mg/L and a mixture of volatile fatty acids was fed to three hybrid reactors. The fourth reactor, control system, was fed with a wastewater containing sodium acetate and mineral nutrients. Coal waste water contained phenol (490 mg/L); m-, o-, p-cresols (123.0, 58.6, 42 mg/L); 2,4-, 2,5-, 3,4- and 3,5-dimethyl phenols (6.3, 6.3, 4.4 and 21.3mg/L) as major phenolic compounds. A mixture of anaerobic digester sludge and partially granulated sludge (3:1) were used as seed materials for the start up of the reactors. Granules were observed after 45 days of operation of the systems. The granules ranged from 0.4 to 1.2 mm in diameter with good settling characteristics with an SVI of 12 mL/gSS. After granulation, the hybrid reactor performed steadily with phenolics and COD removal efficiencies of 93% and 88%, respectively at volumetric loading rate of 2.24 g COD/Ld and hydraulic retention time of 24 h. The removal efficiencies for phenol and m/p-cresols reached 92% and 93% (corresponding to 450.8 and 153 mg/L), while o-cresol was degraded to 88% (corresponding to 51.04 mg/L). Dimethyl phenols could be removed completely at all the organic loadings and did not contribute much to the residual organics. Biodegradation of o-cresol was obtained in the hybrid-UASB reactors. PMID:16762495

  5. A systems analysis of the ARIES tokamak reactors

    SciTech Connect

    Bathke, C.G.

    1992-10-01

    The multi-institutional ARIES study has completed a series of cost-of-electricity optimized conceptual designs of commercial tokamak fusion reactors that vary the assumed advances in technology and physics. A comparison of these designs indicates the cost benefit of various design options. A parametric systems analysis suggests a possible means to obtain a marginally competitive fusion reactor.

  6. A systems analysis of the ARIES tokamak reactors

    SciTech Connect

    Bathke, C.G.

    1992-01-01

    The multi-institutional ARIES study has completed a series of cost-of-electricity optimized conceptual designs of commercial tokamak fusion reactors that vary the assumed advances in technology and physics. A comparison of these designs indicates the cost benefit of various design options. A parametric systems analysis suggests a possible means to obtain a marginally competitive fusion reactor.

  7. Liquid Lithium Divertor System in a Spherical Tokamak Reactor

    Microsoft Academic Search

    Yoshio Nagayama

    2009-01-01

    The heat flux in the divertor target is one of the most crucial problems for a fusion reactor. This problem is more severe in a spherical tokamak (ST) reactor because it is more compact. This paper proposes a liquid lithium divertor system to solve this problem. The heat coming from the fusion plasma along the divertor leg is removed by

  8. SP-100 Program: space reactor system and subsystem investigations

    SciTech Connect

    Harty, R.B.

    1983-09-30

    For a space reactor power system, a comprehensive safety program will be required to assure that no undue risk is present. This report summarizes the nuclear safety review/approval process that will be required for a space reactor system. The documentation requirements are presented along with a summary of the required contents of key documents. Finally, the aerospace safety program conducted for the SNAP-10A reactor system is summarized. The results of this program are presented to show the type of program that can be expected and to provide information that could be usable in future programs.

  9. Gas-cooled reactor for space power systems

    SciTech Connect

    Walter, C.E.; Pearson, J.S.

    1987-05-01

    Reactor characteristics based on extensive development work on the 500-MWt reactor for the Pluto nuclear ramjet are described for space power systems useful in the range of 2 to 20 MWe for operating times of 1 y. The modest pressure drop through the prismatic ceramic core is supported at the outlet end by a ceramic dome which also serves as a neutron reflector. Three core materials are considered which are useful at temperatures up to about 2000 K. Most of the calculations are based on a beryllium oxide with uranium dioxide core. Reactor control is accomplished by use of a burnable poison, a variable-leakage reflector, and internal control rods. Reactivity swings of 20% are obtained with a dozen internal boron-10 rods for the size cores studied. Criticality calculations were performed using the ALICE Monte Carlo code. The inherent high-temperature capability of the reactor design removes the reactor as a limiting condition on system performance. The low fuel inventories required, particularly for beryllium oxide reactors, make space power systems based on gas-cooled near-thermal reactors a lesser safeguard risk than those based on fast reactors.

  10. Robotic system for remote maintenance of a pulsed nuclear reactor

    SciTech Connect

    Thunborg, S.

    1986-01-01

    Guidelines recently established for occupational radiation exposure specify that exposure should be as low as reasonably achievable. In conformance with these guidelines, SNL has developed a remote maintenance robot (RMR) system for use in the Sandia Pulse Reactor III (SPR III) facility. The RMR should reduce occupational radiation exposure by a factor of 4 and decrease reactor downtime. Other goals include developing a technology base for a more advanced pulse reactor and for the nuclear fuel cycle programs of the US Department of Energy and US Nuclear Regulatory Commission. The RMR has five major subsystems: (a) a chain-driven cart to bring the system into the reactor room; (b) a Puma 560 robot to perform dextrous operations; (c) a programmable turntable to orient the robot to any of the reactor's four sides; (d) a programmable overhead hoist for lifting components weighing up to 400 lb onto or off of the reactor; and (e) a supervisory control console for the system operator. Figure 1 is a schematic diagram of the turntable, hoist, and robot system in position around the SPR III reactor.

  11. Microprocessor tester for the treat upgrade reactor trip system

    SciTech Connect

    Lenkszus, F.R.; Bucher, R.G.

    1984-01-01

    The upgrading of the Transient Reactor Test (TREAT) Facility at ANL-Idaho has been designed to provide additional experimental capabilities for the study of core disruptive accident (CDA) phenomena. In addition, a programmable Automated Reactor Control System (ARCS) will permit high-power transients up to 11,000 MW having a controlled reactor period of from 15 to 0.1 sec. These modifications to the core neutronics will improve simulation of LMFBR accident conditions. Finally, a sophisticated, multiply-redundant safety system, the Reactor Trip System (RTS), will provide safe operation for both steady state and transient production operating modes. To insure that this complex safety system is functioning properly, a Dedicated Microprocessor Tester (DMT) has been implemented to perform a thorough checkout of the RTS prior to all TREAT operations.

  12. Behaviour of pharmaceuticals and endocrine disrupting chemicals in simplified sewage treatment systems.

    PubMed

    Brandt, Emanuel M F; de Queiroz, Fernanda B; Afonso, Robson J C F; Aquino, Sérgio F; Chernicharo, Carlos A L

    2013-10-15

    This work assessed the behaviour of nine pharmaceuticals and/or endocrine disrupting chemicals (EDCs) in demo-scale upflow anaerobic sludge blanket reactors (UASB reactors) coupled to distinct simplified post-treatment units (submerged bed, polishing ponds, and trickling filters) fed on raw sewage taken from a municipality in Brazil. The dissolved concentration of the studied micropollutants in the raw and treated sewage was obtained using solid phase extraction (SPE) followed by analysis in a liquid chromatography system coupled to a hybrid high resolution mass spectrometer consisting of an ion-trap and time of flight (LC-MS-IT-TOF). The UASB reactors demonstrated that they were not appropriate for efficiently removing the assessed compounds from the sewage. Furthermore, this study demonstrated that the hydraulic retention time (HRT) was an important parameter for the removal of the hydrophilic and less biodegradable compounds, such as trimethoprim and sulfamethoxazole. The post-treatment units substantially increased the removal of most target micropollutants present in the anaerobic effluents, with a greater removal of micropollutants in simplified systems that require a large construction area, such as the submerged bed and polishing ponds, probably because of the higher HRT employed. Alternatively, compact post-treatment systems, such as trickling filters, tended to be less effective at removing most of the micropollutants studied, and the type of packing proved to be crucial for determining the fate of such compounds using trickling filters. PMID:23850766

  13. System pressure effect on the nuclear reactor limiting criterion

    SciTech Connect

    Chen, Kuo-Fu.

    1990-01-01

    The acceptable operating limits of a nuclear reactor are set to prevent fuel cladding damage. Critical Heat Flux (CHF) is the limiting criterion for the high pressure systems such as the BWRs (6.9 MPa) and the PWRs (13.8 MPa). However, the Onset of Flow Instability (OFI) is the limiting criterion of the low pressure system such as the existing Savannah River Site (SRS) production reactors (0.2 MPa). The physical basis of this difference is presented. 3 refs.

  14. Digital, remote control system for a 2-MW research reactor

    SciTech Connect

    Battle, R.E.; Corbett, G.K.

    1988-01-01

    A fault-tolerant programmable logic controller (PLC) and operator workstations have been programmed to replace the hard-wired relay control system in the 2-MW Bulk Shielding Reactor (BSR) at Oak Ridge National Laboratory. In addition to the PLC and remote and local operator workstations, auxiliary systems for remote operation include a video system, an intercom system, and a fiber optic communication system. The remote control station, located at the High Flux Isotope Reactor 2.5 km from the BSR, has the capability of rector startup and power control. The system was designed with reliability and fail-safe features as important considerations. 4 refs., 3 figs.

  15. Passive cooling system for nuclear reactor containment structure

    DOEpatents

    Gou, Perng-Fei (Saratoga, CA); Wade, Gentry E. (Saratoga, CA)

    1989-01-01

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

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

    DOEpatents

    Gou, Perng-Fei (Saratoga, CA); Wade, Gentry E. (Saratoga, CA)

    1990-01-01

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

  17. Proposal of Space Reactor for Nuclear Electric Propulsion System

    NASA Astrophysics Data System (ADS)

    Nagata, Hidetaka; Nishiyama, Takaaki; Nakashima, Hideki

    Currently, the solar battery, the chemical cell, and the RI battery are used for the energy source in space. However, it is difficult for them to satisfy requirements for deep space explorations. Therefore, other electric power sources which can stably produce high electric energy output, regardless of distance from the sun, are necessary to execute such missions. Then, we here propose small nuclear reactors as power sources for deep space exploration, and consider a conceptual design of a small nuclear reactor for Nuclear Electric Propulsion System. It is found from nuclear analyses that the Gas-Cooled reactor could not meet the design requirement imposed on the core mass. On the other hand, a light water reactor is found to be a promising alternative to the Gas-Cooled reactor.

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

    SciTech Connect

    Kambe, M. (Komae Research Laboratory, Central Research Institute of Electric Power Industry (CRIEPI) 11-1, Iwato Kita 2-chome, Komae-shi, Tokyo 201 (Japan))

    1993-01-15

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

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

    SciTech Connect

    Poston, David I.; Kapernick, Richard J.; Dixon, David D.; Amiri, Benjamin W.; Marcille, Thomas F. [Nuclear Systems Design Group, Los Alamos National Laboratory, Los Alamos, New Mexico, 87545 (United States)

    2008-01-21

    Surface fission power systems on the Moon and Mars may provide the first US application of fission reactor technology in space since 1965. The requirements of many surface power applications allow the consideration of systems with much less development risk than most other space reactor applications, because of modest power (10s of kWe) and no driving need for minimal mass (allowing temperatures <1000 K). The Affordable Fission Surface Power System (AFSPS) study was completed by NASA/DOE to determine the cost of a modest performance, low-technical risk surface power system. This paper describes the reference AFSPS reactor module concept, which is designed to provide a net power of 40 kWe for 8 years on the lunar surface; note, the system has been designed with technologies that are fully compatible with a Martian surface application. The reactor concept uses stainless-steel based, UO{sub 2}-fueled, liquid metal-cooled fission reactor coupled to free-piston Stirling converters. The reactor shielding approach utilizes both in-situ and launched shielding to keep the dose to astronauts much lower than the natural background radiation on the lunar surface. One of the important 'affordability' attributes is that the concept has been designed to minimize both the technical and programmatic safety risk.

  20. Computer based systems in boiling water reactors

    Microsoft Academic Search

    J. N. Shukla; J. A. Iubelt

    1980-01-01

    This paper describes the application of computers to the General Electric Company's Boilling Water Reactor (BWR) type nuclear power plants. In the GE BWR plants, computers are used for Real Time Process Monitoring, Nuclear Steam Supply Systemn Performance and Core Limit Evaluation, Balance of Plant Performance Evaluation, Historical Recording, and Control Rod Pattern Enforcement. These functions are performed by different

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

    SciTech Connect

    Bloomfield, H.S.

    1987-12-01

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

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

    NASA Technical Reports Server (NTRS)

    Bloomfield, Harvey S.

    1987-01-01

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

  3. Reactor Lithium Heat Pipes for HP-STMCs Space Reactor Power System

    SciTech Connect

    Tournier, Jean-Michel; El-Genk, Mohamed S. [Institute for Space and Nuclear Power Studies, University of New Mexico, Albuquerque, NM, 87131 (United States); Chemical and Nuclear Engineering Dept., The University of New Mexico, Albuquerque, NM, 87131 (United States)

    2004-02-04

    Design and performance analysis of the nuclear reactor's lithium heat pipes for a 110-kWe Heat Pipes-Segmented Thermoelectric Module Converters (HP-STMCs) Space Reactor Power system (SRPS) are presented. The evaporator length of the heat pipes is the same as the active core height (0.45 m) and the C-C finned condenser is of the same length as the STMC panels (1.5 m). The C-C finned condenser section is radiatively coupled to the collector shoes of the STMCs placed on both sides. The lengths of the adiabatic section, the values of the power throughput and the evaporator wall temperature depend on the radial location of the heat pipe in the reactor core and the number and dimensions of the potassium heat pipes in the heat rejection radiator. The reactor heat pipes have a total length that varies from 7.57 to 7.73 m, and a 0.2 mm thick Mo-14%Re wick with an average pore radius of 12 {mu}m. The wick is separated from the Mo-14%Re wall by a 0.5 mm annulus filled with liquid lithium, to raise the prevailing capillary limit. The nominal evaporator (or reactor) temperature varies from 1513 to 1591 K and the thermal power of the reactor is 1.6 MW, which averages 12.7 kW for each of the 126 reactor heat pipes. The power throughput per heat pipe increase to a nominal 15.24 kW at the location of the peak power in the core and to 20.31 kW when an adjacent heat pipe fails. The prevailing capillary limit of the reactor heat pipes is 28.3 kW, providing a design margin {>=} 28%.

  4. Microbial dynamics in upflow anaerobic sludge blanket (UASB) bioreactor granules in response to short-term changes in substrate feed

    SciTech Connect

    Kovacik, William P.; Scholten, Johannes C.; Culley, David E.; Hickey, Robert; Zhang, Weiwen; Brockman, Fred J.

    2010-08-01

    The complexity and diversity of the microbial communities in biogranules from an upflow anaerobic sludge blanket (UASB) bioreactor were determined in response to short-term changes in substrate feeds. The reactor was fed simulated brewery wastewater (SBWW) (70% ethanol, 15% acetate, 15% propionate) for 1.5 months (phase 1), acetate / sulfate for 2 months (phase 2), acetate-alone for 3 months (phase 3), and then a return to SBWW for 2 months (phase 4). Performance of the reactor remained relatively stable throughout the experiment as shown by COD removal and gas production. 16S rDNA, methanogen-associated mcrA and sulfate reducer-associated dsrAB genes were PCR amplified, then cloned and sequenced. Sequence analysis of 16S clone libraries showed a relatively simple community composed mainly of the methanogenic Archaea (Methanobacterium and Methanosaeta), members of the Green Non-Sulfur (Chloroflexi) group of Bacteria, followed by fewer numbers of Syntrophobacter, Spirochaeta, Acidobacteria and Cytophaga-related Bacterial sequences. Methanogen-related mcrA clone libraries were dominated throughout by Methanobacter and Methanospirillum related sequences. Although not numerous enough to be detected in our 16S rDNA libraries, sulfate reducers were detected in dsrAB clone libraries, with sequences related to Desulfovibrio and Desulfomonile. Community diversity levels (Shannon-Weiner index) generally decreased for all libraries in response to a change from SBWW to acetate-alone feed. But there was a large transitory increase noted in 16S diversity at the two-month sampling on acetate-alone, entirely related to an increase in Bacterial diversity. Upon return to SBWW conditions in phase 4, all diversity measures returned to near phase 1 levels.

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

    PubMed

    Corradini, M L

    2007-11-01

    Almost 450 nuclear power plants are currently operating throughout the world and supplying about 17% of the world's electricity. These plants perform safely, reliably, and have no free-release of byproducts to the environment. Given the current rate of growth in electricity demand and the ever growing concerns for the environment, nuclear power can only satisfy the need for electricity and other energy-intensive products if it can demonstrate (1) enhanced safety and system reliability, (2) minimal environmental impact via sustainable system designs, and (3) competitive economics. The U.S. Department of Energy with the international community has begun research on the next generation of nuclear energy systems that can be made available to the market by 2030 or earlier, and that can offer significant advances toward these challenging goals; in particular, six candidate reactor system designs have been identified. These future nuclear power systems will require advances in materials, reactor physics, as well as thermal-hydraulics to realize their full potential. However, all of these designs must demonstrate enhanced safety above and beyond current light water reactor systems if the next generation of nuclear power plants is to grow in number far beyond the current population. This paper reviews the advanced Generation-IV reactor systems and the key safety phenomena that must be considered to guarantee that enhanced safety can be assured in future nuclear reactor systems. PMID:18049233

  6. Architecture of the ETR (experimental test reactor) systems code

    SciTech Connect

    Reid, R.L.; Galambos, J.D.

    1987-01-01

    TETRA, a tokamak systems code capable of modeling experimental test reactors (ETRs), was developed in a joint effort by participants of the fusion community. The first version of this code was constructed to model devices similar to the Tokamak Ignition/Burn Engineering Reactor (TIBER) in configuration and design. A major feature of this code is its ability to perform optimization studies. Future work will include broadening the scope of the code, particularly in the area of materials selection, to more accurately simulate tokamak configurations such as the Next European Torus (NET) and the Fusion Engineering Reactor (FER). 18 refs., 2 figs., 4 tabs.

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

    NASA Astrophysics Data System (ADS)

    El-Genk, Mohamed S.

    2009-05-01

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

  8. Dual annular rotating "windowed" nuclear reflector reactor control system

    DOEpatents

    Jacox, Michael G. (Idaho Falls, ID); Drexler, Robert L. (Idaho Falls, ID); Hunt, Robert N. M. (Idaho Falls, ID); Lake, James A. (Idaho Falls, ID)

    1994-01-01

    A nuclear reactor control system is provided in a nuclear reactor having a core operating in the fast neutron energy spectrum where criticality control is achieved by neutron leakage. The control system includes dual annular, rotatable reflector rings. There are two reflector rings: an inner reflector ring and an outer reflector ring. The reflectors are concentrically assembled, surround the reactor core, and each reflector ring includes a plurality of openings. The openings in each ring are capable of being aligned or non-aligned with each other. Independent driving means for each of the annular reflector rings is provided so that reactor criticality can be initiated and controlled by rotation of either reflector ring such that the extent of alignment of the openings in each ring controls the reflection of neutrons from the core.

  9. System aspects of a Space Nuclear Reactor Power System

    SciTech Connect

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

    1988-01-01

    Selected systems aspects of a 300 kW nuclear reactor power system for spacecraft have been studied. The approach included examination of two candidate missions and their associated spacecraft, and a number of special topics dealing with the power system design and operation. The missions considered were a reusable orbital transfer vehicle and a space-based radar. The special topics included: power system configuration and scaling, launch vehicle integration, operating altitude, orbital storage, start-up, thawing, control, load following, procedures in case of malfunction, restart, thermal and nuclear radiation to other portions of the spacecraft, thermal stresses between subsystems, boom and cable designs, vibration modes, altitude control, reliability, and survivability. Among the findings are that the stowed length of the power system is important to mission design and that orbital storage for months to years may be needed for missions involving orbital assembly. The power system design evolved during the study and has continued to evolve; the current design differs somewhat from that examined in this paper.

  10. Advanced Fuel Cycle Economic Analysis of Symbiotic Light-Water Reactor and Fast Burner Reactor Systems

    SciTech Connect

    D. E. Shropshire

    2009-01-01

    The Advanced Fuel Cycle Economic Analysis of Symbiotic Light-Water Reactor and Fast Burner Reactor Systems, prepared to support the U.S. Advanced Fuel Cycle Initiative (AFCI) systems analysis, provides a technology-oriented baseline system cost comparison between the open fuel cycle and closed fuel cycle systems. The intent is to understand their overall cost trends, cost sensitivities, and trade-offs. This analysis also improves the AFCI Program’s understanding of the cost drivers that will determine nuclear power’s cost competitiveness vis-a-vis other baseload generation systems. The common reactor-related costs consist of capital, operating, and decontamination and decommissioning costs. Fuel cycle costs include front-end (pre-irradiation) and back-end (post-iradiation) costs, as well as costs specifically associated with fuel recycling. This analysis reveals that there are large cost uncertainties associated with all the fuel cycle strategies, and that overall systems (reactor plus fuel cycle) using a closed fuel cycle are about 10% more expensive in terms of electricity generation cost than open cycle systems. The study concludes that further U.S. and joint international-based design studies are needed to reduce the cost uncertainties with respect to fast reactor, fuel separation and fabrication, and waste disposition. The results of this work can help provide insight to the cost-related factors and conditions needed to keep nuclear energy (including closed fuel cycles) economically competitive in the U.S. and worldwide. These results may be updated over time based on new cost information, revised assumptions, and feedback received from additional reviews.

  11. Pressurized reactor system and a method of operating the same

    DOEpatents

    Isaksson, Juhani M. (Karhula, FI)

    1996-01-01

    A method and apparatus are provided for operating a pressurized reactor system in order to precisely control the temperature within a pressure vessel in order to minimize condensation of corrosive materials from gases on the surfaces of the pressure vessel or contained circulating fluidized bed reactor, and to prevent the temperature of the components from reaching a detrimentally high level, while at the same time allowing quick heating of the pressure vessel interior volume during start-up. Superatmospheric pressure gas is introduced from the first conduit into the fluidized bed reactor and heat derived reactions such as combustion and gassification are maintained in the reactor. Gas is exhausted from the reactor and pressure vessel through a second conduit. Gas is circulated from one part of the inside volume to another to control the temperature of the inside volume, such as by passing the gas through an exterior conduit which has a heat exchanger, control valve, blower and compressor associated therewith, or by causing natural convection flow of circulating gas within one or more generally vertically extending gas passages entirely within the pressure vessel (and containing heat exchangers, flow rate control valves, or the like therein). Preferably, inert gas is provided as a circulating gas, and the inert gas may also be used in emergency shut-down situations. In emergency shut-down reaction gas being supplied to the reactor is cut off, while inert gas from the interior gas volume of the pressure vessel is introduced into the reactor.

  12. Nuclear reactor heat transport system component low friction support system

    DOEpatents

    Wade, Elman E. (Ruffs Dale, PA)

    1980-01-01

    A support column for a heavy component of a liquid metal fast breeder reactor heat transport system which will deflect when the pipes leading coolant to and from the heavy component expand or contract due to temperature changes includes a vertically disposed pipe, the pipe being connected to the heavy component by two longitudinally spaced cycloidal dovetail joints wherein the distal end of each of the dovetails constitutes a part of the surface of a large diameter cylinder and the centerlines of these large diameter cylinders intersect at right angles and the pipe being supported through two longitudinally spaced cycloidal dovetail joints wherein the distal end of each of the dovetails constitutes a part of the surface of a large diameter cylinder and the centerlines of these large diameter cylinders intersect at right angles, each of the cylindrical surfaces bearing on a flat and horizontal surface.

  13. Software reliability and safety in nuclear reactor protection systems

    SciTech Connect

    Lawrence, J.D. [Lawrence Livermore National Lab., CA (United States)

    1993-11-01

    Planning the development, use and regulation of computer systems in nuclear reactor protection systems in such a way as to enhance reliability and safety is a complex issue. This report is one of a series of reports from the Computer Safety and Reliability Group, Lawrence Livermore that investigates different aspects of computer software in reactor National Laboratory, that investigates different aspects of computer software in reactor protection systems. There are two central themes in the report, First, software considerations cannot be fully understood in isolation from computer hardware and application considerations. Second, the process of engineering reliability and safety into a computer system requires activities to be carried out throughout the software life cycle. The report discusses the many activities that can be carried out during the software life cycle to improve the safety and reliability of the resulting product. The viewpoint is primarily that of the assessor, or auditor.

  14. Westinghouse Reactor Protection System Unavailability, 1984-1995

    SciTech Connect

    C. D. Gentillon; D. Marksberry (USNRC); D. Rasmuson; M. B. Calley; S. A. Eide; T. Wierman (INEEL)

    1999-08-01

    An analysis was performed of the safety-related performance of the reactor protection system (RPS) at U.S. Westinghouse commercial reactors during the period 1984 through 1995. RPS operational data were collected from the Nuclear Plant Reliability Data System and Licensee Event Reports. A risk-based analysis was performed on the data to estimate the observed unavailability of the RPS, based on a fault tree model of the system. Results were compared with existing unavailability estimates from Individual Plant Examinations and other reports.

  15. Different Mechanisms for Establishing Liquid Walls in Advanced Reactor Systems

    NASA Astrophysics Data System (ADS)

    Hançerlio?ullari, Aybaba; Cini, Mesut

    2013-04-01

    The APEX study is investigating the use of free flowing liquid surfaces to form the inner surface of the chamber around a fusion plasma. In this study the modeling of APEX hybrid reactor produced by using ARIES-RS hybrid reactor technology, was performed by using the Monte Carlo code and ENF/B-V-VI nuclear data. The most important feature of APEX hybrid reactor is that the first wall surrounding the plasma is liquid. The advantages of utilizing a liquid wall are high power density capacity, good power transformation productivity the magnitude of the reactor's operational duration, low failure percentage, short maintenance time and the inclusion of the system's simple technology and material. Around the fusion chamber, molten salt Li2BeF4 and natural lithium were used as cooling materials. The result of the study indicated that fissile material production UF4 and ThF4 heavy metal salt increased nearly at the same percentage.

  16. Reactor Controllability of 3-Region-Core Molten Salt Reactor System - A Study on Load Following Capability

    SciTech Connect

    Takahisa Yamamoto; Koshi Mitachi; Masatoshi Nishio [Toyohashi University of Technology, 1-1, Hibarigaoka, Tempaku-cho, Toyohashi-shi Aichi, 4418580 (Japan)

    2006-07-01

    The Molten Salt Reactor (MSR) systems are liquid-fueled reactors that can be used for actinide burning, production of electricity, production of hydrogen, and production of fissile fuels (breeding). Thorium (Th) and uranium-233 ({sup 233}U) are fertile and fissile of the MSR systems, and dissolved in a high-temperature molten fluoride salt (fuel salt) with a very high boiling temperature (up to 1650 K), that is both the reactor nuclear fuel and the coolant. The MSR system is one of the six advanced reactor concepts identified by the Generation IV International Forum (GIF) as a candidate for cooperative development. In the MSR system, fuel salt flows through a fuel duct constructed around a reactor core and fuel channel of a graphite moderator accompanied by fission reaction and heat generation, and flows out to an external-loop system consisted of a heat exchanger and a circulation pump. Due to the motion of fuel salt, delayed neutron precursors that are one of the source of neutron production make to change their position between the fission reaction and neutron emission events and decay even occur in the external loop system. Hence the reactivity and effective delayed neutron precursor fraction of the MSR system are lower than those of solid fuel reactor systems such as Boiling Water Reactors (BWRs) and Pressurised Water Reactor (PWRs). Since all of the presently operating nuclear power reactors utilize solid fuel, little attention had been paid to the MSR analysis of the reactivity loss and reactor characteristics change caused by the fuel salt circulation. Sides et al. and Shimazu et al. developed MSR analytical models based on the point reactor kinetics model to consider the effect of fuel salt flow. Their models represented a reactor as having six zones for fuel salt and three zones for the graphite moderator. Since their models employed the point reactor kinetics model and the rough temperature approximation, their results were not sufficiently accurate to consider the effect of fuel salt flow. (authors)

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

    SciTech Connect

    Josephson, Gary B.; Tonkyn, Russell G.; Frye, J. G.; Riley, Brian J.; Rappe, Kenneth G.

    2011-04-06

    Pacific Northwest National Laboratory (PNNL) has performed an assessment of a Hybrid Plasma/Filter system as an alternative to conventional methods for collective protection. The key premise of the hybrid system is to couple a nonthermal plasma (NTP) reactor with reactive adsorption to provide a broader envelope of protection than can be provided through a single-solution approach. The first step uses highly reactive species (e.g. oxygen radicals, hydroxyl radicals, etc.) created in a nonthermal plasma (NTP) reactor to destroy the majority (~75% - 90%) of an incoming threat. Following the NTP reactor an O3 reactor/filter uses the O3 created in the NTP reactor to further destroy the remaining organic materials. This report summarizes the laboratory development of the Hybrid Plasma Reactor/Filter to protect against a ‘worst-case’ simulant, methyl bromide (CH3Br), and presents a preliminary engineering assessment of the technology to Joint Expeditionary Collective Protection performance specifications for chemical vapor air purification technologies.

  18. Reactor technology assessment and selection utilizing systems engineering approach

    SciTech Connect

    Zolkaffly, Muhammed Zulfakar; Han, Ki-In [KEPCO International Nuclear Graduate School, Ulsan (Korea, Republic of)

    2014-02-12

    The first Nuclear power plant (NPP) deployment in a country is a complex process that needs to consider technical, economic and financial aspects along with other aspects like public acceptance. Increased interest in the deployment of new NPPs, both among newcomer countries and those with expanding programs, necessitates the selection of reactor technology among commercially available technologies. This paper reviews the Systems Decision Process (SDP) of Systems Engineering and applies it in selecting the most appropriate reactor technology for the deployment in Malaysia. The integrated qualitative and quantitative analyses employed in the SDP are explored to perform reactor technology assessment and to select the most feasible technology whose design has also to comply with the IAEA standard requirements and other relevant requirements that have been established in this study. A quick Malaysian case study result suggests that the country reside with PWR (pressurized water reactor) technologies with more detailed study to be performed in the future for the selection of the most appropriate reactor technology for Malaysia. The demonstrated technology assessment also proposes an alternative method to systematically and quantitatively select the most appropriate reactor technology.

  19. Reactor technology assessment and selection utilizing systems engineering approach

    NASA Astrophysics Data System (ADS)

    Zolkaffly, Muhammed Zulfakar; Han, Ki-In

    2014-02-01

    The first Nuclear power plant (NPP) deployment in a country is a complex process that needs to consider technical, economic and financial aspects along with other aspects like public acceptance. Increased interest in the deployment of new NPPs, both among newcomer countries and those with expanding programs, necessitates the selection of reactor technology among commercially available technologies. This paper reviews the Systems Decision Process (SDP) of Systems Engineering and applies it in selecting the most appropriate reactor technology for the deployment in Malaysia. The integrated qualitative and quantitative analyses employed in the SDP are explored to perform reactor technology assessment and to select the most feasible technology whose design has also to comply with the IAEA standard requirements and other relevant requirements that have been established in this study. A quick Malaysian case study result suggests that the country reside with PWR (pressurized water reactor) technologies with more detailed study to be performed in the future for the selection of the most appropriate reactor technology for Malaysia. The demonstrated technology assessment also proposes an alternative method to systematically and quantitatively select the most appropriate reactor technology.

  20. STORS: BATTELLE-NORTHWEST'S SLUDGE TO OIL REACTOR SYSTEM

    EPA Science Inventory

    A continuous primary sewage sludge thermochemical conversion system was designed, built, and operated for over 100 hr during 1984 at Battelle-Northwest. This was designated as the STORS (Sludge-to-Oil-Reactor System), because the major product was a burnable oil. Starting with 20...

  1. The University of Missouri Research Reactor facility can melter system

    Microsoft Academic Search

    C. B. Jr. Edwards; O. L. Olson; R. Stevens; R. M. Brugger

    1987-01-01

    At the University of Missouri Research Reactor (MURR), a waste compacting system for reducing the volume of radioactive aluminum cans has been designed, built and put into operation. In MURR's programs of producing radioisotopes and transmutation doping of silicon, a large volume of radioactive aluminum cans is generated. The Can Melter System (CMS) consists of a sorting station, a can

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

    DOEpatents

    Gaubatz, D.C.

    1996-12-17

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

  3. Small reactor power systems for manned planetary surface bases

    SciTech Connect

    Bloomfield, H.S.

    1987-12-01

    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.

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

    DOEpatents

    Gaubatz, Donald C. (Cupertino, CA)

    1996-01-01

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

  5. Small reactor power systems for manned planetary surface bases

    NASA Technical Reports Server (NTRS)

    Bloomfield, Harvey S.

    1987-01-01

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

  6. A helium heat-removal system for reactor subsystem testing

    NASA Astrophysics Data System (ADS)

    Zabriskie, J. M.; Landman, W. H., Jr.

    This paper describes a secondary heat-removal system for proposed ground-based SP-100 Space Nuclear Power Reactor testing. This system uses helium gas to remove heat from the primary coolant and rejects that heat to a water system. The system consist of four circulators, a primary liquid-metal/helium heat exchanger, two helium/water heat exchangers, and various flow or temperature control valves and instrumentation. Much of this system is currently installed at the deactivated Engineering Test Reactor (ETR) at the Idaho National Engineering Laboratory (INEL). Modifications are described which would allow higher heat removal capacities. An innovative remote-joint design using a phase-change material is presented as a possible solution to the problem of joining the reactor test module to the facility helium circulator system while inside a vacuum chamber. Also presented is a proposed design for a liquid-metal/helium heat exchanger using low cost materials, shapes, and fabrication processes to interface between the SP-100 space reactor heat pipe wells and the helium heat-removal system.

  7. System Study: Reactor Core Isolation Cooling 1998–2012

    SciTech Connect

    T. E. Wierman

    2013-10-01

    This report presents an unreliability evaluation of the reactor core isolation cooling (RCIC) system at 31 U.S. commercial boiling water reactors. Demand, run hours, and failure data from fiscal year 1998 through 2012 for selected components were obtained from the Equipment Performance and Information Exchange (EPIX). The unreliability results are trended for the most recent 10 year period while yearly estimates for system unreliability are provided for the entire active period. No statistically significant increasing trend was identified in the HPCI results. Statistically significant decreasing trends were identified for RCIC start-only and 8-hour trends.

  8. Comparison of conventional and membrane reactor fuel processors for hydrocarbon-based PEM fuel cell systems

    Microsoft Academic Search

    James R. Lattner; Michael P. Harold

    2004-01-01

    Several reactor types for the autothermal reforming (ATR) of hydrocarbon fuels are evaluated for the production of hydrogen in PEM fuel cell systems. Each ATR reactor is integrated into an overall process model including the fuel cell, heat integration exchangers, and water recycle. Hydrogen permselective membrane reactors (Pd-based and proton conducting oxide) are compared to the three-step reactor system consisting

  9. Designs for remote inspection of the ALMR Reactor Vessel Auxiliary Cooling System (RVACS)

    Microsoft Academic Search

    F. J. Sweeney; D. G. Carroll; C. Chen; C. Crane; R. Dalton; J. R. Taylor; S. Tosunoglu; T. Weymouth

    1993-01-01

    One of the most important safety systems in General Electric`s (GI) Advanced Liquid Metal Reactor (ALMR) is the Reactor Vessel Auxiliary Cooling System (RVACS). Because of high temperature, radiation, and restricted space conditions, GI desired methods to remotely inspect the RVACS, emissive coatings, and reactor vessel welds during normal refueling operations. The DOE\\/NE Robotics for Advanced Reactors program formed a

  10. Designs for remote inspection of the ALMR Reactor Vessel Auxiliary Cooling System (RVACS)

    Microsoft Academic Search

    F. J. Sweeney; D. G. Carroll; C. Chen; C. Crane; R. Dalton; J. R. Taylor; S. Tosunoglu

    1993-01-01

    One of the most important safety systems in General Electric's (GI) Advanced Liquid Metal Reactor (ALMR) is the Reactor Vessel Auxiliary Cooling System (RVACS). Because of high temperature, radiation, and restricted space conditions, GI desired methods to remotely inspect the RVACS, emissive coatings, and reactor vessel welds during normal refueling operations. The DOE\\/NE Robotics for Advanced Reactors program formed a

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

    DOEpatents

    Hunsbedt, Anstein (Los Gatos, CA); Busboom, Herbert J. (San Jose, CA)

    1991-01-01

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

  12. Incipient Transient Detection in Reactor Systems: Experimental and Theoretical Investigation

    SciTech Connect

    Lefteri H. Tsoukalas; S.T. Revankar; X Wang; R. Sattuluri

    2005-09-27

    The main goal of this research was to develop a method for detecting reactor system transients at the earliest possible time through a comprehensive experimental, testing and benchmarking program. This approach holds strong promise for developing new diagnostic technologies that are non-intrusive, generic and highly portable across different systems. It will help in the design of new generation nuclear power reactors, which utilize passive safety systems with a reliable and non-intrusive multiphase flow diagnostic system to monitor the function of the passive safety systems. The main objective of this research was to develop an improved fuzzy logic based detection method based on a comprehensive experimental testing program to detect reactor transients at the earliest possible time, practically at their birth moment. A fuzzy logic and neural network based transient identification methodology and implemented in a computer code called PROTREN was considered in this research and was compared with SPRT (Sequentially Probability Ratio Testing) decision and Bayesian inference. The project involved experiment, theoretical modeling and a thermal-hydraulic code assessment. It involved graduate and undergraduate students participation providing them with exposure and training in advanced reactor concepts and safety systems. In this final report, main tasks performed during the project period are summarized and the selected results are presented. Detailed descriptions for the tasks and the results are presented in previous yearly reports (Revankar et al 2003 and Revankar et al 2004).

  13. Design, analysis and optimization of the power conversion system for the Modular Pebble Bed Reactor System

    E-print Network

    Wang, Chunyun, 1968-

    2003-01-01

    The Modular Pebble Bed Reactor system (MPBR) requires a gas turbine cycle (Brayton cycle) as the power conversion system for it to achieve economic competitiveness as a GenIV nuclear system. The availability of controllable ...

  14. Effect of influent COD/SO4(2-) ratios on UASB treatment of a synthetic sulfate-containing wastewater.

    PubMed

    Hu, Yong; Jing, Zhaoqian; Sudo, Yuta; Niu, Qigui; Du, Jingru; Wu, Jiang; Li, Yu-You

    2015-07-01

    The effect of the chemical oxygen demand/sulfate (COD/SO4(2-)) ratio on the anaerobic treatment of synthetic chemical wastewater containing acetate, ethanol, and sulfate, was investigated using a UASB reactor. The experimental results show that at a COD/SO4(2-) ratio of 20 and a COD loading rate of 25.2gCODL(-1)d(-1), a COD removal of as high as 87.8% was maintained. At a COD/SO4(2-) ratio of 0.5 (sulfate concentration 6000mgL(-1)), however, the COD removal was 79.2% and the methane yield was 0.20LCH4gCOD(-1). The conversion of influent COD to methane dropped from 80.5% to 54.4% as the COD/SO4(2-) ratio decreased from 20 to 0.5. At all the COD/SO4(2-) ratios applied, over 79.4% of the total electron flow was utilized by methane-producing archaea (MPA), indicating that methane fermentation was the predominant reaction. The majority of the methane was produced by acetoclastic MPA at high COD/SO4(2-) ratios and both acetoclastic and hydrogenthrophic MPA at low COD/SO4(2-) ratios. Only at low COD/SO4(2-) ratios were SRB species such as Desulfovibrio found to play a key role in ethanol degradation, whereas all the SRB species were found to be incomplete oxidizers at both high and low COD/SO4(2-) ratios. PMID:25747303

  15. Deployment of remote dismantlement systems at the CP-5 reactor

    SciTech Connect

    Black, D.B.; Ditch, R.W.; Henley, D.R.; Seifert, L.S.

    1997-06-01

    The Chicago Pile 5 (CP-5) Reactor Facility is currently undergoing decontamination and decommissioning (D&D) at the Argonne National Laboratory (ANL) Illinois site. CP-5 was the principal nuclear reactor used to produce neutrons for scientific research at Argonne from 1954 to 1979. The CP-5 reactor was a heavy-water moderated, enriched uranium-fueled reactor with a graphite reflector. The CP-5 D&D project includes the disassembly and removal of all radioactive components, equipment, and structures associated with the CP-5 facility. The Department of Energy`s Robotics Technology Development Program along with the Federal Energy Technology Center, Morgantown Office, have provided teleoperated, remote systems for use in the dismantlement of the CP-5 reactor structure for tasks requiring remote dismantlement. These systems include the dual-arm work platform, the Rosie mobile D&D vehicle, the swing-reduced crane control system, and a remotely-operated crane control system. The dual-arm work platform is a robotic dismantlement system that includes a pair of Schilling Titan III hydraulic manipulators mounted on a special platform, a hydraulic power unit and an operator console. The Rosie mobile D&D work system developed by RedZone Robotics, Inc. is an electro-hydraulic omni-directional locomotor platform with a heavy manipulator mounted on its deck. The Rosie vehicle moves about the floor around the CP-5 reactor block and is operated from a console in the control room. The swing-reduced crane control system has been installed on the CP-5 polar crane, and allows a load suspended from the crane hook to be moved while reducing the induced swing in the load. A remote control system and a rotating crane hook have also been added to the CP-5 polar crane. This paper discusses the status of these remote systems at CP-5 and the facility changes made to allow for their use in the dismantlement of the reactor structure internals. 4 refs., 3 figs.

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

    SciTech Connect

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

    2006-04-01

    This report describes the results of a Sandia National Laboratories internally funded research program to study the coupling of nuclear reactors to gas dynamic Brayton power conversion systems. The research focused on developing integrated dynamic system models, fabricating a 10-30 kWe closed loop Brayton cycle, and validating these models by operating the Brayton test-loop. The work tasks were performed in three major areas. First, the system equations and dynamic models for reactors and Closed Brayton Cycle (CBC) systems were developed and implemented in SIMULINKTM. Within this effort, both steady state and dynamic system models for all the components (turbines, compressors, reactors, ducting, alternators, heat exchangers, and space based radiators) were developed and assembled into complete systems for gas cooled reactors, liquid metal reactors, and electrically heated simulators. Various control modules that use proportional-integral-differential (PID) feedback loops for the reactor and the power-conversion shaft speed were also developed and implemented. The simulation code is called RPCSIM (Reactor Power and Control Simulator). In the second task an open cycle commercially available Capstone C30 micro-turbine power generator was modified to provide a small inexpensive closed Brayton cycle test loop called the Sandia Brayton test-Loop (SBL-30). The Capstone gas-turbine unit housing was modified to permit the attachment of an electrical heater and a water cooled chiller to form a closed loop. The Capstone turbine, compressor, and alternator were used without modification. The Capstone systems nominal operating point is 1150 K turbine inlet temperature at 96,000 rpm. The annular recuperator and portions of the Capstone control system (inverter) and starter system also were reused. The rotational speed of the turbo-machinery is controlled by adjusting the alternator load by using the electrical grid as the load bank. The SBL-30 test loop was operated at the manufacturers site (Barber-Nichols Inc.) and installed and operated at Sandia. A sufficiently detailed description of the loop is provided in this report along with the design characteristics of the turbo-alternator-compressor set to allow other researchers to compare their results with those measured in the Sandia test-loop. The third task consisted of a validation effort. In this task the test loop was operated and compared with the modeled results to develop a more complete understanding of this electrically heated closed power generation system and to validate the model. The measured and predicted system temperatures and pressures are in good agreement, indicating that the model is a reasonable representation of the test loop. Typical deviations between the model and the hardware results are less than 10%. Additional tests were performed to assess the capability of the Brayton engine to continue to remove decay heat after the reactor/heater is shutdown, to develop safe and effective control strategies, and to access the effectiveness of gas inventory control as an alternative means to provide load following. In one test the heater power was turned off to simulate a rapid reactor shutdown, and the turbomachinery was driven solely by the sensible heat stored in the heater for over 71 minutes without external power input. This is an important safety feature for CBC systems as it means that the closed Brayton loop will keep cooling the reactor without the need for auxiliary power (other than that needed to circulate the waste heat rejection coolant) provided the heat sink is available.

  17. STORS: THE SLUDGE-TO-OIL REACTOR SYSTEM

    EPA Science Inventory

    The Sludge-to-Oil Reactor System (STORS) continuously converted over 400 gallons (20 percent solids) of sewage sludge to oil during 100 hours of operation. About 80 percent of the energy in the sludge was recovered as an oil and char. The energy recovered was sufficient to make t...

  18. CRAC2. Code System for Calculation of Reactor Accident Consequences

    Microsoft Academic Search

    L. T. Ritchie; J. D. Johnson; R. M. Blond

    1983-01-01

    The CRAC code system was developed in support of the Reactor Safety Study (WASH-1400) to access the risk from potential accidents at nuclear power plants. CRAC2 was developed to satisfy the need for more realistic consequence estimation techniques to be used for such purposes as site evaluation, emergency planning and response, and general risk assessment and to correct errors which

  19. Ensuring required reliability for nuclear reactor protection systems

    Microsoft Academic Search

    A. I. Pereguda; A. A. Petrenko

    1989-01-01

    One of the fundamental questions arising in nuclear plant design is ensuring safe operation in various operating regimes. Safe reactor operation is achieved not only by ensuring high values for the equipment reliability indicators during normal operation, but also by the correct actions of plant staff and of the safety systems. Since the failure of various types of equipment can

  20. Space-reactor electric systems: subsystem technology assessment

    SciTech Connect

    Anderson, R.V.; Bost, D.; Determan, W.R.

    1983-03-29

    This report documents the subsystem technology assessment. For the purpose of this report, five subsystems were defined for a space reactor electric system, and the report is organized around these subsystems: reactor; shielding; primary heat transport; power conversion and processing; and heat rejection. The purpose of the assessment was to determine the current technology status and the technology potentials for different types of the five subsystems. The cost and schedule needed to develop these potentials were estimated, and sets of development-compatible subsystems were identified.

  1. A Gas-Cooled Reactor Surface Power System

    SciTech Connect

    Harms, G.A.; Lenard, R.X.; Lipinski, R.J.; Wright, S.A.

    1998-11-09

    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 nitide clad in Nb 1 %Zr, which has been extensively tested under the SP-I 00 program The fiel rods are arranged in a hexagonal array within a BeO block. The BeO softens the spectrum, allowing better use of the fbel and stabilizing the geometty 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 cannot occur for any launch accident. This system is also adaptable for electric propulsion or life-support during transit to and from Mars.

  2. Ultra-reliable computer systems: an integrated approach for application in reactor safety systems

    SciTech Connect

    Chisholm, G.H.

    1985-01-01

    Improvements in operation and maintenance of nuclear reactors can be realized with the application of computers in the reactor control systems. In the context of this paper a reactor control system encompasses the control aspects of the Reactor Safety System (RSS). Equipment qualification for application in reactor safety systems requires a rigorous demonstration of reliability. For the purpose of this paper, the reliability demonstration will be divided into two categories. These categories are demonstrations of compliance with respect to (a) environmental; and (b) functional design constrains. This paper presents an approach for the determination of computer-based RSS respective to functional design constraints only. It is herein postulated that the design for compliance with environmental design constraints is a reasonably definitive problem and within the realm of available technology. The demonstration of compliance with design constraints respective to functionality, as described herein, is an extension of available technology and requires development.

  3. Expert system for online surveillance of nuclear reactor coolant pumps

    DOEpatents

    Gross, Kenny C. (Bolingbrook, IL); Singer, Ralph M. (Naperville, IL); Humenik, Keith E. (Columbia, MD)

    1993-01-01

    An expert system for online surveillance of nuclear reactor coolant pumps. This system provides a means for early detection of pump or sensor degradation. Degradation is determined through the use of a statistical analysis technique, sequential probability ratio test, applied to information from several sensors which are responsive to differing physical parameters. The results of sequential testing of the data provide the operator with an early warning of possible sensor or pump failure.

  4. Neutral-beam systems for magnetic-fusion reactors

    SciTech Connect

    Fink, J. H.

    1981-08-10

    Neutral beams for magnetic fusion reactors are at an early stage of development, and require considerable effort to make them into the large, reliable, and efficient systems needed for future power plants. To optimize their performance to establish specific goals for component development, systematic analysis of the beamlines is essential. Three ion source characteristics are discussed: arc-cathode life, gas efficiency, and beam divergence, and their significance in a high-energy neutral-beam system is evaluated.

  5. Operation of staged membrane oxidation reactor systems

    DOEpatents

    Repasky, John Michael

    2012-10-16

    A method of operating a multi-stage ion transport membrane oxidation system. The method comprises providing a multi-stage ion transport membrane oxidation system with at least a first membrane oxidation stage and a second membrane oxidation stage, operating the ion transport membrane oxidation system at operating conditions including a characteristic temperature of the first membrane oxidation stage and a characteristic temperature of the second membrane oxidation stage; and controlling the production capacity and/or the product quality by changing the characteristic temperature of the first membrane oxidation stage and/or changing the characteristic temperature of the second membrane oxidation stage.

  6. Granules characteristics in the vertical profile of a full-scale upflow anaerobic sludge blanket reactor treating poultry slaughterhouse wastewater.

    PubMed

    Del Nery, Valéria; Pozzi, Eloisa; Damianovic, Márcia H R Z; Domingues, Mércia R; Zaiat, Marcelo

    2008-04-01

    The performance and the granules characteristics of a 450 m(3) -UASB reactor operating for 1228 days, treating poultry slaughterhouse wastewater with an average COD reduction of 85% was examined. Granules were sampled in three different positions along the vertical central line of the reactor, revealing variations in the concentration of volatile total solids. Although the reactor had been in operation for an extended period of time, granule sizes of 0.5-1.5 mm appeared to predominate. The hollow core was well defined for granules with sizes ranging from 2 to 3 mm in all the sampling ports. The granules exhibited no layered microbial distribution and were packed with different morphotype cells intertwined randomly throughout the cross-section. Methanogenic Archaea predominated in the granules taken from every sampling port along the reactor. The results indicated that the characterization of the granules is a useful tool for the adoption of operational strategies toward optimization of UASB reactors. PMID:17478089

  7. 78 FR 41436 - Proposed Revision to Treatment of Non-Safety Systems for Passive Advanced Light Water Reactors

    Federal Register 2010, 2011, 2012, 2013, 2014

    2013-07-10

    ...Non-Safety Systems for Passive Advanced Light Water Reactors AGENCY: Nuclear Regulatory...Systems (RTNSS) for Passive Advanced Light Water Reactors.'' The NRC seeks public...Systems (RTNSS) for Passive Advanced Light Water Reactors.'' This area...

  8. A passive automated personnel accountability system for reactor emergency preparedness

    SciTech Connect

    Zimmerman, R.O.; DeLisle, G.V.; Hickey, E.E.

    1988-04-01

    In 1985 a project was undertaken at the N Reactor on the Hanford Site to develop an automated personnel accountability system to ensure accountability of all personnel within 30 minutes of a site evacuation. The decision to develop such a system was made after a full-scale evacuation drill showed that the manual accountability system in use at the time was inadequate to meet the 30-minute requirement. Accountability systems at commercial nuclear power plants were evaluated, but found to be unsuitable because they were not passive, that is, they required action on part of the user for the system to work. Approximately 2500 people could be required to evacuate the 100-N Area. Therefore, a card key system or badge exchange system was judged not to be feasible. A passive accountability system was desired for N Reactor to allow personnel to enter and leave the site in a more timely manner. To meet the need for an automated accountability system at N Reactor, a special Evacuation Accountability System (EVACS) was designed and developed. The EVACS system has three basic components: the transponder, a credit card-sized device worn with the security badge; portal monitors, which are electronically activated by the transponder; and a computer information system that contains the personnel data base. Each person wearing a transponder is accounted for automatically by walking through a portal. In this paper, a description of the hardware and software will be presented, together with problems encountered and lessons learned while adapting an existing technology to this particular use. The system is currently installed and requires acceptance testing before becoming operational.

  9. Biofilm reactors for industrial bioconversion processes: employing potential of enhanced reaction rates

    PubMed Central

    Qureshi, Nasib; Annous, Bassam A; Ezeji, Thaddeus C; Karcher, Patrick; Maddox, Ian S

    2005-01-01

    This article describes the use of biofilm reactors for the production of various chemicals by fermentation and wastewater treatment. Biofilm formation is a natural process where microbial cells attach to the support (adsorbent) or form flocs/aggregates (also called granules) without use of chemicals and form thick layers of cells known as "biofilms." As a result of biofilm formation, cell densities in the reactor increase and cell concentrations as high as 74 gL-1 can be achieved. The reactor configurations can be as simple as a batch reactor, continuous stirred tank reactor (CSTR), packed bed reactor (PBR), fluidized bed reactor (FBR), airlift reactor (ALR), upflow anaerobic sludge blanket (UASB) reactor, or any other suitable configuration. In UASB granular biofilm particles are used. This article demonstrates that reactor productivities in these reactors have been superior to any other reactor types. This article describes production of ethanol, butanol, lactic acid, acetic acid/vinegar, succinic acid, and fumaric acid in addition to wastewater treatment in the biofilm reactors. As the title suggests, biofilm reactors have high potential to be employed in biotechnology/bioconversion industry for viable economic reasons. In this article, various reactor types have been compared for the above bioconversion processes. PMID:16122390

  10. 77 FR 15812 - Initial Test Program of Condensate and Feedwater Systems for Light-Water Reactors

    Federal Register 2010, 2011, 2012, 2013, 2014

    2012-03-16

    ...of Condensate and Feedwater Systems for Light-Water Reactors AGENCY: Nuclear Regulatory...of Condensate and Feedwater Systems for Light- Water Reactors.'' DG-1265 is proposed...condensate and feedwater systems in all types of light water reactor facilities licensed...

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

    SciTech Connect

    Buden, D.

    1993-08-11

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

  12. Advanced ultrasonic inspection system upgrades for reactor vessel

    SciTech Connect

    Taniguchi, M.; Aoyama, T.; Yoshioka, K. [Mitsubishi Heavy Industries, Ltd., Kobe (Japan); Omichi, T. [Mitsubishi Heavy Industries, Ltd., Takasago (Japan)

    1995-08-01

    Ultrasonic inspection systems consisting of large and stiff columns and guide rails have been usually used for inservice inspections of reactor vessels (RV) of PWR nuclear power plants. However, these are so large and heavy that handling and maintenance are very complicated. To solve these problems, the authors developed a prototype advanced ultrasonic inspection system (A-UT system) which was a submarine type, and applied to a preservice inspection in August 1991. Based on the experience, the authors have improved the system for inservice inspections in the following points: (1) Reliability of system; (2) Laser positioning system; (3) Control system; and (4) Data acquisition system, etc. It is expected to reduce critical path time of RV-ISI dramatically using this system.

  13. Common cause analysis of the TREAT upgrade reactor protection system

    SciTech Connect

    Page, R.J.; Kamis, G.J.; Marbach, R.A.; Mueller, C.J.

    1984-09-01

    A triply redundant reactor scram system (RSS) has been designed for the upgraded TREAT facility. The independent failures reliability goal for the RSS is <10/sup -9/ failures per demand. An independent failures analysis indicated that this goal would be met. In addition, however, recognizing that in heavily redundant systems common-cause failures dominate, a common cause analysis of the TREAT upgrade RSS was done. The objective was to identify those common-cause initiators which could affect the functioning of the RSS, and to subsequently modify the design of the RSS so that the effect was minimized. A number of common-cause initiators were identified which were capable of defeating the triple redundancy feature of the reactor scram system. By means of a systematic analysis of the effect these initiators could have on the system, it was possible to identify seven necessary design and procedural modifications that would greatly reduce the probability of the reactor being run while the RSS was in a faulted condition.

  14. Monitoring nuclear reactor systems using neural networks and fuzzy logic

    SciTech Connect

    Ikonomopoulos, A.; Tsoukalas, L.H.; Uhrig, R.E. [Tennessee Univ., Knoxville, TN (United States); Mullens, J.A. [Tennessee Univ., Knoxville, TN (United States)]|[Oak Ridge National Lab., TN (United States)

    1991-12-01

    A new approach is presented that demonstrates the potential of trained artificial neural networks (ANNs) as generators of membership functions for the purpose of monitoring nuclear reactor systems. ANN`s provide a complex-to-simple mapping of reactor parameters in a process analogous to that of measurement. Through such ``virtual measurements`` the value of parameters with operational significance, e.g., control-valve-disk-position, valve-line-up or performance can be determined. In the methodology presented the output of a virtual measuring device is a set of membership functions which independently represent different states of the system. Utilizing a fuzzy logic representation offers the advantage of describing the state of the system in a condensed form, developed through linguistic descriptions and convenient for application in monitoring, diagnostics and generally control algorithms. The developed methodology is applied to the problem of measuring the disk position of the secondary flow control valve of an experimental reactor using data obtained during a start-up. The enhanced noise tolerance of the methodology is clearly demonstrated as well as a method for selecting the actual output. The results suggest that it is possible to construct virtual measuring devices through artificial neural networks mapping dynamic time series to a set of membership functions and thus enhance the capability of monitoring systems. 8 refs., 11 figs., 1 tab.

  15. Monitoring system for a liquid-cooled nuclear fission reactor

    DOEpatents

    DeVolpi, Alexander (Bolingbrook, IL)

    1987-01-01

    A monitoring system for detecting changes in the liquid levels in various regions of a water-cooled nuclear power reactor, viz., in the downcomer, in the core, in the inlet and outlet plenums, at the head, and elsewhere; and also for detecting changes in the density of the liquid in these regions. A plurality of gamma radiation detectors are used, arranged vertically along the outside of the reactor vessel, and collimator means for each detector limits the gamma-radiation it receives as emitting from only isolated regions of the vessel. Excess neutrons produced by the fission reaction will be captured by the water coolant, by the steel reactor walls, or by the fuel or control structures in the vessel. Neutron capture by steel generates gamma radiation having an energy level of the order of 5-12 MeV, whereas neutron capture by water provides an energy level of approximately 2.2 MeV, and neutron capture by the fission fuel or its cladding provides an energy level of 1 MeV or less. The intensity of neutron capture thus changes significantly at any water-metal interface. Comparative analysis of adjacent gamma detectors senses changes from the normal condition with liquid coolant present to advise of changes in the presence and/or density of the coolant at these specific regions. The gamma detectors can also sense fission-product gas accumulation at the reactor head to advise of a failure of fuel-pin cladding.

  16. Systems and methods for dismantling a nuclear reactor

    DOEpatents

    Heim, Robert R; Adams, Scott Ryan; Cole, Matthew Denver; Kirby, William E; Linnebur, Paul Damon

    2014-10-28

    Systems and methods for dismantling a nuclear reactor are described. In one aspect the system includes a remotely controlled heavy manipulator ("manipulator") operatively coupled to a support structure, and a control station in a non-contaminated portion of a workspace. The support structure provides the manipulator with top down access into a bioshield of a nuclear reactor. At least one computing device in the control station provides remote control to perform operations including: (a) dismantling, using the manipulator, a graphite moderator, concrete walls, and a ceiling of the bioshield, the manipulator being provided with automated access to all internal portions of the bioshield; (b) loading, using the manipulator, contaminated graphite blocks from the graphite core and other components from the bioshield into one or more waste containers; and (c) dispersing, using the manipulator, dust suppression and contamination fixing spray to contaminated matter.

  17. Post impact behavior of mobile reactor core containment systems

    NASA Technical Reports Server (NTRS)

    Puthoff, R. L.; Parker, W. G.; Vanbibber, L. E.

    1972-01-01

    The reactor core containment vessel temperatures after impact, and the design variables that affect the post impact survival of the system are analyzed. The heat transfer analysis includes conduction, radiation, and convection in addition to the core material heats of fusion and vaporization under partially burial conditions. Also, included is the fact that fission products vaporize and transport radially outward and condense outward and condense on cooler surfaces, resulting in a moving heat source. A computer program entitled Executive Subroutines for Afterheat Temperature Analysis (ESATA) was written to consider this complex heat transfer analysis. Seven cases were calculated of a reactor power system capable of delivering up to 300 MW of thermal power to a nuclear airplane.

  18. Designing visual displays and system models for safe reactor operations

    SciTech Connect

    Brown-VanHoozer, S.A.

    1995-12-31

    The material presented in this paper is based on two studies involving the design of visual displays and the user`s prospective model of a system. The studies involve a methodology known as Neuro-Linguistic Programming and its use in expanding design choices from the operator`s perspective image. The contents of this paper focuses on the studies and how they are applicable to the safety of operating reactors.

  19. 77 FR 62270 - Proposed Revision Treatment of Non-Safety Systems for Passive Advanced Light Water Reactors

    Federal Register 2010, 2011, 2012, 2013, 2014

    2012-10-12

    ...Non-Safety Systems for Passive Advanced Light Water Reactors AGENCY: Nuclear Regulatory...Systems (RTNSS) for Passive Advanced Light Water Reactors.'' The current SRP does...the proposed RTNSS for Passive Advance Light Water Reactors. DATES: Submit...

  20. Space-reactor system and subsystem investigations: Cost and schedule estimates for reactor and shield subsystems technology development. SP100 program

    Microsoft Academic Search

    W. R. Determan; R. B. Harty; C. Hylin

    1983-01-01

    Cost and schedule estimates of the technology development for reactor and shielding subsystems of a 100-kWe class space reactor electric system are presented. The subsystems technology development (which includes reactor and shield subsystems ground testing) is supported by materials and processes development and component development. For the purpose of the cost estimate, seven generic types of reactor subsystems were used:

  1. 77 FR 55877 - Initial Test Program of Condensate and Feedwater Systems for Light-Water Reactors

    Federal Register 2010, 2011, 2012, 2013, 2014

    2012-09-11

    ...of Condensate and Feedwater Systems for Light-Water Reactors AGENCY: Nuclear Regulatory...condensate and feedwater systems in all types of light water reactor facilities; and (2) to...including condensate storage and supply, for light-water reactors (LWRs) and for...

  2. Risk assessment of computer-controlled safety systems for fusion reactors

    Microsoft Academic Search

    M. O. Fryer; S. Z. Bruske

    1983-01-01

    The complexity of fusion reactor systems and the need to display, analyze, and react promptly to large amounts of information during reactor operation will require a number of safety systems in the fusion facilities to be computer controlled. Computer software, therefore, must be included in the reactor safety analyses. Unfortunately, the science of integrating computer software into safety analyses is

  3. Reactor Controllability of 3-Region-Core Molten Salt Reactor System - A Study on Load Following Capability

    Microsoft Academic Search

    Takahisa Yamamoto; Koshi Mitachi; Masatoshi Nishio

    2006-01-01

    The Molten Salt Reactor (MSR) systems are liquid-fueled reactors that can be used for actinide burning, production of electricity, production of hydrogen, and production of fissile fuels (breeding). Thorium (Th) and uranium-233 (²³³U) are fertile and fissile of the MSR systems, and dissolved in a high-temperature molten fluoride salt (fuel salt) with a very high boiling temperature (up to 1650

  4. Small-scale domestic wastewater treatment using an alternating pumped sequencing batch biofilm reactor system.

    PubMed

    Rodgers, Michael; Zhan, Xinmin; O'Reilly, Edmond

    2006-04-01

    An alternating pumped sequencing batch biofilm reactor (APSBBR) system was developed to treat small-scale domestic wastewater. This laboratory system had two reactor tanks, Reactor 1 and Reactor 2, with two identical plastic biofilm modules in each reactor. Reactor 1 of the APSBBR had five operational phases--fill, anoxic, aerobic, settle and draw. In the aerobic phase, the wastewater was circulated between the two reactor tanks with centrifugal pumps and aeration was mainly achieved through oxygen absorption by microorganisms in the biofilms when they were exposed to the air. This paper details the performance of the APSBBR system in treating synthetic domestic wastewater over 18 months. The effluent from the APSBBR system satisfied the European Wastewater Treatment Directive requirements, with respect to COD, ammonium-nitrogen and suspended solids. The biofilm growth in the two reactor tanks was different due to the difference in substrate loadings and growth conditions. PMID:16408190

  5. Evaluation of an anaerobic\\/aerobic system for carbon and nitrogen removal in slaughterhouse wastewater

    Microsoft Academic Search

    L. A. Núñez; B. Martínez

    2001-01-01

    In this work the performance of an anaerobic UASB reactor coupled with an activated sludge reactor for carbon and nitrogen removal in slaughterhouse wastewater is investigated. Periods with and without recirculation of aerobic effluent over 165 days are analysed. Working with a recirculation ratio of 2, removal efficiencies up to 90% and 65% are obtained for DQO and total nitrogen

  6. Water gas shift membrane reactor for CO 2 control in IGCC systems: techno-economic feasibility study

    Microsoft Academic Search

    M Bracht; P. T Alderliesten; R Kloster; R Pruschek; G Haupt; E Xue; J. R. H Ross; M. K Koukou; N Papayannakos

    1997-01-01

    A novel reactor concept, the water gas shift membrane reactor (WGS-MR) for CO2 removal in IGCC systems has been investigated. In order to establish full insight in the possibilities of the application of such a reactor, a multidisciplinary feasibility study has been carried out comprising system integration studies, catalyst research, membrane research, membrane reactor modelling and bench scale membrane reactor

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

    NASA Technical Reports Server (NTRS)

    Lantz, E.; Mayo, W.

    1972-01-01

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

  8. Concept of magnet systems for LHD-type reactor

    NASA Astrophysics Data System (ADS)

    Imagawa, S.; Takahata, K.; Tamura, H.; Yanagi, N.; Mito, T.; Obana, T.; Sagara, A.

    2009-07-01

    Heliotron reactors have attractive features for fusion power plants such as having no need for current drive and a wide space between the helical coils for the maintenance of in-vessel components. Their main disadvantage was considered to be the necessarily large size of their magnet systems. According to the recent reactor studies based on the experimental results in the Large Helical Device, a major radius of plasma of 14-17 m with a central toroidal field of 6-4 T is needed to attain the self-ignition condition with a blanket space thicker than 1.1 m. The stored magnetic energy is estimated at 120-140 GJ. Although both the major radius and the magnetic energy are about three times as large as ITER, the maximum magnetic field and mechanical stress are comparable. In the preliminary structural analysis, the maximum stress intensity including the peak stress is less than the 1000 MPa that is allowed for strengthened stainless steel. Although the length of the helical coil is more than 150 m, that is about five times as long as the ITER TF coil, cable-in-conduit conductors can be adopted with a parallel winding method of five-in-hand. The concept of the parallel winding is proposed. Consequently, the magnet systems for helical reactors can be realized with a small extension of the ITER technology.

  9. Autonomous Control Capabilities for Space Reactor Power Systems

    NASA Astrophysics Data System (ADS)

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

    2004-02-01

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

  10. Magnet systems for the International Thermonuclear Experimental Reactor

    SciTech Connect

    Henning, C.D.; Miller, J.R.

    1988-09-22

    The definition phase for the International Thermonuclear Experimental Reactor (ITER) has been nearly completed, thus beginning a three-year design effort by teams from the European Community (EC), Japan, US, and USSR. Preliminary parameters for the superconducting magnet system have been established to guide more detailed design work. Radiation tolerance of the superconductors and insulators has been important because it sets requirements for the neutron-shield dimension and sensitively influences reactor size. Major levels of mechanical stress appear in the structural cases of the inboard legs of the toroidal-field (TF) coils. The winding packs of the TF coils include significant fractions of steel that provide support against in-plane separating loads, but they offer little support against out-of-plane loads unless shear-bonding of the conductors can be maintained. Heat removal from nuclear and ac loads has not limited the fundamental design, but it has nonnegligible economic consequences. 3 refs., 3 figs., 5 tabs.

  11. Pressurized water nuclear reactor system with hot leg vortex mitigator

    DOEpatents

    Lau, Louis K. S. (Monroeville, PA)

    1990-01-01

    A pressurized water nuclear reactor system includes a vortex mitigator in the form of a cylindrical conduit between the hot leg conduit and a first section of residual heat removal conduit, which conduit leads to a pump and a second section of residual heat removal conduit leading back to the reactor pressure vessel. The cylindrical conduit is of such a size that where the hot leg has an inner diameter D.sub.1, the first section has an inner diameter D.sub.2, and the cylindrical conduit or step nozzle has a length L and an inner diameter of D.sub.3 ; D.sub.3 /D.sub.1 is at least 0.55, D.sub.2 is at least 1.9, and L/D.sub.3 is at least 1.44, whereby cavitation of the pump by a vortex formed in the hot leg is prevented.

  12. Ongoing Development of a Series Bosch Reactor System

    NASA Technical Reports Server (NTRS)

    Abney, Morgan; Mansell, Matt; DuMez, Sam; Thomas, John; Cooper, Charlie; Long, David

    2013-01-01

    Future manned missions to deep space or planetary surfaces will undoubtedly require highly robust, efficient, and regenerable life support systems that require minimal consumables. To meet this requirement, NASA continues to explore a Bosch-based carbon dioxide reduction system to recover oxygen from CO2. In order to improve the equivalent system mass of Bosch systems, we seek to design and test a "Series Bosch" system in which two reactors in series are optimized for the two steps of the reaction, as well as to explore the use of in situ materials as carbon deposition catalysts. Here we report recent developments in this effort including assembly and initial testing of a Reverse Water-Gas Shift reactor (RWGSr) and initial testing of two gas separation membranes. The RWGSr was sized to reduce CO2 produced by a crew of four to carbon monoxide as the first stage in a Series Bosch system. The gas separation membranes, necessary to recycle unreacted hydrogen and CO2, were similarly sized. Additionally, we report results of preliminary experiments designed to determine the catalytic properties of Martian and Lunar regolith simulant for the carbon deposition step.

  13. Ongoing Development of a Series Bosch Reactor System

    NASA Technical Reports Server (NTRS)

    Abney, Morgan B; Mansell, J. Matthew; Stanley, Christine; Edmunson, Jennifer; DuMez, Samuel J.; Chen, Kevin

    2013-01-01

    Future manned missions to deep space or planetary surfaces will undoubtedly incorporate highly robust, efficient, and regenerable life support systems that require minimal consumables. To meet this requirement, NASA continues to explore a Bosch-based carbon dioxide reduction system to recover oxygen from CO2. In order to improve the equivalent system mass of Bosch systems, we seek to design and test a "Series Bosch" system in which two reactors in series are optimized for the two steps of the reaction, as well as to explore the use of in situ materials as carbon deposition catalysts. Here we report recent developments in this effort including assembly and initial testing of a Reverse Water-Gas Shift reactor (RWGSr) and initial testing of two gas separation membranes. The RWGSr was sized to reduce CO2 produced by a crew of four to carbon monoxide as the first stage in a Series Bosch system. The gas separation membranes, necessary to recycle unreacted hydrogen and CO2, were similarly sized. Additionally, we report results of preliminary experiments designed to determine the catalytic properties of Martian regolith simulant for the carbon formation step.

  14. Nuclear safety as applied to space power reactor systems

    SciTech Connect

    Cummings, G.E.

    1987-01-01

    To develop a strategy for incorporating and demonstrating safety, it is necessary to enumerate the unique aspects of space power reactor systems from a safety standpoint. These features must be differentiated from terrestrial nuclear power plants so that our experience can be applied properly. Some ideas can then be developed on how safe designs can be achieved so that they are safe and perceived to be safe by the public. These ideas include operating only after achieving a stable orbit, developing an inherently safe design, ''designing'' in safety from the start and managing the system development (design) so that it is perceived safe. These and other ideas are explored further in this paper.

  15. System for fuel rod removal from a reactor module

    DOEpatents

    Matchett, Richard L. (Bethel Park, PA); Roof, David R. (North Huntingdon, PA); Kikta, Thomas J. (Pittsburgh, PA); Wilczynski, Rosemarie (McKees Rocks, PA); Nilsen, Roy J. (Pittsburgh, PA); Bacvinskas, William S. (Bethel Park, PA); Fodor, George (Pittsburgh, PA)

    1990-01-01

    A robotic system for remote underwater withdrawal of the fuel rods from fuel modules of a light water breeder reactor includes a collet/grapple assembly for gripping and removing fuel rods in each module, which is positioned by use of a winch and a radial support means attached to a vertical support tube which is mounted over the fuel module. A programmable logic controller in conjunction with a microcomputer, provides control for the accurate positioning and pulling force of the rod grapple assembly. Closed circuit television cameras are provided which aid in operator interface with the robotic system.

  16. System for fuel rod removal from a reactor module

    DOEpatents

    Matchett, R.L.; Fodor, G.; Kikta, T.J.; Bacvinsicas, W.S.; Roof, D.R.; Nilsen, R.J.; Wilczynski, R.

    1988-07-28

    A robotic system for remote underwater withdrawal of the fuel rods from fuel modules of a light water breeder reactor includes a collet/grapple assembly for gripping and removing fuel rods in each module, which is positioned by use of a winch and a radial support means attached to a vertical support tube which is mounted over the fuel module. A programmable logic controller in conjunction with a microcomputer, provides control for the accurate positioning and pulling force of the rod grapple assembly. Closed circuit television cameras are provided which aid in operator interface with the robotic system. 7 figs.

  17. Nuclear plant-aging research on reactor protection systems

    SciTech Connect

    Meyer, L.C.

    1988-01-01

    This report presents the rsults of a review of the Reactor Trip System (RTS) and the Engineered Safety Feature Actuating System (ESFAS) operating experiences reported in Licensee Event Reports (LER)s, the Nuclear Power Experience data base, Nuclear Plant Reliability Data System, and plant maintenance records. Our purpose is to evaluate the potential significance of aging, including cycling, trips, and testing as contributors to degradation of the RTS and ESFAS. Tables are presented that show the percentage of events for RTS and ESFAS classified by cause, components, and subcomponents for each of the Nuclear Steam Supply System vendors. A representative Babcock and Wilcox plant was selected for detailed study. The US Nuclear Regulatory Commission's Nuclear Plant Aging Research guidelines were followed in performing the detailed study that identified materials susceptible to aging, stressors, environmental factors, and failure modes for the RTS and ESFAS as generic instrumentation and control systems. Functional indicators of degradation are listed, testing requirements evaluated, and regulatory issues discussed.

  18. Aging study of boiling water reactor high pressure injection systems

    SciTech Connect

    Conley, D.A.; Edson, J.L.; Fineman, C.F. [Lockheed Idaho Technologies Co., Idaho Falls, ID (United States)

    1995-03-01

    The purpose of high pressure injection systems is to maintain an adequate coolant level in reactor pressure vessels, so that the fuel cladding temperature does not exceed 1,200{degrees}C (2,200{degrees}F), and to permit plant shutdown during a variety of design basis loss-of-coolant accidents. This report presents the results of a study on aging performed for high pressure injection systems of boiling water reactor plants in the United States. The purpose of the study was to identify and evaluate the effects of aging and the effectiveness of testing and maintenance in detecting and mitigating aging degradation. Guidelines from the United States Nuclear Regulatory Commission`s Nuclear Plant Aging Research Program were used in performing the aging study. Review and analysis of the failures reported in databases such as Nuclear Power Experience, Licensee Event Reports, and the Nuclear Plant Reliability Data System, along with plant-specific maintenance records databases, are included in this report to provide the information required to identify aging stressors, failure modes, and failure causes. Several probabilistic risk assessments were reviewed to identify risk-significant components in high pressure injection systems. Testing, maintenance, specific safety issues, and codes and standards are also discussed.

  19. Characteristics of Spent Fuel from Plutonium Disposition Reactors, Vol. 1: The Combustion Engineering System 80+ Pressurized-Water-Reactor Design

    Microsoft Academic Search

    B. D. Murphy

    1993-01-01

    This report discusses a simulation study of the burnup of mixed-oxide fuel in a Combustion Engineering System 80+ Pressurized-Water Reactor. The mixed oxide was composed of uranium and plutonium oxides where the plutonium was of weapons-grade composition. The study was part of the Fissile Materials Disposition Program that considered the possibility of fueling commercial reactors with weapons plutonium. The isotopic

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

    DOEpatents

    Hunsbedt, Anstein (Los Gatos, CA); Boardman, Charles E. (Saratoga, CA)

    1993-01-01

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

  1. Use of lead-bismuth coolant in nuclear reactors and accelerator-driven systems

    Microsoft Academic Search

    B. F. Gromov; Yu. S. Belomitcev; E. I. Yefimov; M. P. Leonchuk; P. N. Martinov; Yu. I. Orlov; D. V. Pankratov; Yu. G. Pashkin; G. I. Toshinsky; V. V. Chekunov; B. A. Shmatko; V. S. Stepanov

    1997-01-01

    Experience of using lead-bismuth coolant in reactors of Russian nuclear submarines is briefly presented. The salient points of the concept providing the safety of reactor facilities cooled by a lead-bismuth eutectic are covered. The key results of developments for use of a lead-bismuth coolant in nuclear reactors and accelerator-driven system, are presented.

  2. Fault-tree analysis of the EBR-II reactor shutdown system

    Microsoft Academic Search

    S. A. Kamal; D. J. Hill

    1991-01-01

    As part of level I Probabilistic Risk Assessment (PRA) of the Experimental Breeder Reactor II (EBR-II), detailed fault trees for the reactor shutdown system are developed. Two classes of transient events that are of particular importance to EBR-II operation and require reactor shutdown are loss of flow (LOF) and transient over power (TOP). Normally in these events, detection channels would

  3. Reliability analysis of the scram system of the Missouri University research reactor

    Microsoft Academic Search

    R. A. Werner; S. K. Loyalka

    1976-01-01

    The reliability analysis of anticipated transients without scram is a topic of considerable significance in reactor safety studies. The article describes the results of a recently completed study on the reliability of the Missouri University Research Reactor (MURR) scram system. For this reactor it has been determined that the failure to initiate a scram automatically or manually within 7.5 sec

  4. ORTAP: a simulator of high temperature gas-cooled reactor nuclear steam supply system dynamics

    Microsoft Academic Search

    J. C. Cleveland; R. A. Hedrick; S. J Bell; J. G. Delene

    1977-01-01

    ORTAP was developed to predict the dynamic behavior of the high temperature gas-cooled reactor (HTGR) Nuclear Steam Supply System for normal operational transients and postulated accidents. Normal operational transients which can be analyzed with ORTAP include both reactor startup and shutdown, and normal and rapid load charges. Upset transients that can be analyzed with ORTAP include reactor trip, turbine trip

  5. Challenges and Innovative Technologies On Fuel Handling Systems for Future Sodium-Cooled Fast Reactors

    E-print Network

    Paris-Sud XI, Université de

    the scheduled refuelling operations. The choice of a FHS impacts directly on the general design of the reactor operating strategies such as core design and management and core configuration. Moreover, the FHS will have handling system (FHS) can be considered as an essential step in the reactor design. The reactor refuelling

  6. 10 CFR 50.46a - Acceptance criteria for reactor coolant system venting systems.

    Code of Federal Regulations, 2011 CFR

    2011-01-01

    ...50.46a Section 50.46a Energy NUCLEAR REGULATORY COMMISSION DOMESTIC LICENSING...coolant system venting systems. Each nuclear power reactor must be provided...required for the tubes in U-tube steam generators. Acceptable venting systems must...

  7. 10 CFR 50.46a - Acceptance criteria for reactor coolant system venting systems.

    Code of Federal Regulations, 2010 CFR

    2010-01-01

    ...50.46a Section 50.46a Energy NUCLEAR REGULATORY COMMISSION DOMESTIC LICENSING...coolant system venting systems. Each nuclear power reactor must be provided...required for the tubes in U-tube steam generators. Acceptable venting systems must...

  8. Method for preventing oxygen corrosion in a boiling water nuclear reactor and improved boiling water reactor system

    SciTech Connect

    Desilva, S.G.

    1989-06-27

    This patent describes a method for prevention of corrosion, due to oxygen activity, in a boiling water nuclear reactor system having a main feedwater recirculation loop comprising a reactor vessel with a reactor core and with a steam off-take line and a condensate return line in flow communication with the vessel for recirculating feedwater between the reactor and a steam turbine. The coolant water recirculation loops consist of coolant water off-take and return lines in flow communication with the vessel for recirculating coolant water in the vessel, wherein hydrogen is added to the coolant water passing through the reactor core to reduce the oxygen content of the coolant water. The improvement consists of: recirculation coolant water containing dissolved oxygen, after passage through the reactor core, is diverted from the recirculation loop; hydrogen gas is added to the diverted coolant water; the hydrogen-containing diverted coolant water is contacted with a catalyst effective to enhance reaction of hydrogen with oxygen contained in the coolant water; and the diverted coolant water, after reaction of the hydrogen gas with the oxygen contained therein, is returned to the recirculation loop for passage through the reactor core.

  9. Technological implications of SNAP reactor power system development on future space nuclear power systems

    SciTech Connect

    Anderson, R.V.

    1982-11-16

    Nuclear reactor systems are one method of satisfying space mission power needs. The development of such systems must proceed on a path consistent with mission needs and schedules. This path, or technology roadmap, starts from the power system technology data base available today. Much of this data base was established during the 1960s and early 1970s, when government and industry developed space nuclear reactor systems for steady-state power and propulsion. One of the largest development programs was the Systems for Nuclear Auxiliary Power (SNAP) Program. By the early 1970s, a technology base had evolved from this program at the system, subsystem, and component levels. There are many implications of this technology base on future reactor power systems. A review of this base highlights the need for performing a power system technology and mission overview study. Such a study is currently being performed by Rockwell's Energy Systems Group for the Department of Energy and will assess power system capabilities versus mission needs, considering development, schedule, and cost implications. The end product of the study will be a technology roadmap to guide reactor power system development.

  10. SP100 space reactor power system for lunar, Mars, and robotic exploration

    Microsoft Academic Search

    Jack F. Mondt

    1992-01-01

    The SP-100 power system is described which was developed for three missions, namely, Pluto Orbiter with nuclear electric propulsion; human-rated surface reactor power system for lunar and Mars exploration; and earth surveillance with an integrated nuclear electric propulsion system. The reactor power systems technology is being developed to meet these requirements so that the technical database, design tools, and specifications

  11. Coil system for a mirror-based hybrid reactor

    SciTech Connect

    Hagnestal, A.; Agren, O.; Moiseenko, V. E. [Uppsala University, Angstroem laboratory, Division of Electricity, Box 534, SE-751 21 Uppsala (Sweden); Institute of Plasma Physics, National Science Center 'Kharkov Institute of Physics and Technology', Akademichna st. 1, 61108 Kharkiv (Ukraine)

    2012-06-19

    Two different superconducting coil systems for the SFLM Hybrid study - a quadrupolar mirror based fusion-fission reactor study - are presented. One coil system is for a magnetic field with 2 T at the midplane and a mirror ratio of four. This coil set consists of semiplanar coils in two layers. The alternative coil system is for a downscaled magnetic field of 1.25 T at the midplane and a mirror ratio of four, where a higher {beta} is required to achieve sufficient the neutron production. This coil set has one layer of twisted 3D coils. The 3D coils are expected to be considerably cheaper than the semiplanar, since NbTi superconductors can be used for most coils instead of Nb3Sn due to the lower magnetic field.

  12. The MAUS nuclear space reactor with ion propulsion system

    NASA Astrophysics Data System (ADS)

    Mainardi, Enrico

    2006-06-01

    MAUS (Moltiplicatore Avanzato Ultracompatto Spaziale) is a nuclear reactor concept design capable to ensure a reliable, long-lasting, low-mass, compact energy supply needed for advanced, future space missions. The exploration of the solar system and the space beyond requires the development of nuclear energy generators for supplying electricity to space-bases, spacecrafts, probes or satellites, as well as for propelling ships in long space missions. For propulsion, the MAUS nuclear reactor could be used to power electric ion drive engines. An ion engine is able to build up to very high velocities, far greater than chemical propulsion systems, but has high power and long service requirements. The MAUS concept is described, together with the ion propulsion engine and together with the reference thermoionic process used to convert the thermal power into electricity. The design work has been performed at the Nuclear Engineering and Energy Conversion Department of the University of Rome "La Sapienza" starting from 1992 on an issue submitted by the Italian Space Agency (ASI), in cooperation with the research laboratories of ENEA.

  13. Nuclear reactor cooling system decontamination reagent regeneration. [PWR; BWR

    DOEpatents

    Anstine, L.D.; James, D.B.; Melaika, E.A.; Peterson, J.P. Jr.

    1980-06-06

    An improved method for decontaminating the coolant system of water-cooled nuclear power reactors and for regenerating the decontamination solution is described. A small amount of one or more weak-acid organic complexing agents is added to the reactor coolant, and the pH is adjusted to form a decontamination solution which is circulated throughout the coolant system to dissolve metal oxides from the interior surfaces and complex the resulting metal ions and radionuclide ions. The coolant containing the complexed metal ions and radionuclide ions is passed through a strong-base anion exchange resin bed which has been presaturated with a solution containing the complexing agents in the same ratio and having the same pH as the decontamination solution. As the decontamination solution passes through the resin bed, metal-complexed anions are exchanged for the metal-ion-free anions on the bed, while metal-ion-free anions in the solution pass through the bed, thus removing the metal ions and regenerating the decontamination solution.

  14. High Flux Isotope Reactor system RELAP5 input model

    SciTech Connect

    Morris, D.G.; Wendel, M.W.

    1993-01-01

    A thermal-hydraulic computational model of the High Flux Isotope Reactor (HFIR) has been developed using the RELAP5 program. The purpose of the model is to provide a state-of-the art thermal-hydraulic simulation tool for analyzing selected hypothetical accident scenarios for a revised HFIR Safety Analysis Report (SAR). The model includes (1) a detailed representation of the reactor core and other vessel components, (2) three heat exchanger/pump cells, (3) pressurizing pumps and letdown valves, and (4) secondary coolant system (with less detail than the primary system). Data from HFIR operation, component tests, tests in facility mockups and the HFIR, HFIR specific experiments, and other pertinent experiments performed independent of HFIR were used to construct the model and validate it to the extent permitted by the data. The detailed version of the model has been used to simulate loss-of-coolant accidents (LOCAs), while the abbreviated version has been developed for the operational transients that allow use of a less detailed nodalization. Analysis of station blackout with core long-term decay heat removal via natural convection has been performed using the core and vessel portions of the detailed model.

  15. A design study on MOX-fueled small fast reactors for standardization of a small fast nuclear reactor system

    Microsoft Academic Search

    Nariaki Uto; Hiroki Hayafune; Toshio Wakabayashi

    2000-01-01

    A way of development to standardize a small fast nuclear reactor system, which is considered one of the suitable concepts at next generation for satisfying such needs as generality, small dependence on natural resources, safety and non-proliferation, is proposed. This process consists of three steps : the first is to demonstrate the basic system within a short period based on

  16. Multi-Purpose Thermal Hydraulic Loop: Advanced Reactor Technology Integral System Test (ARTIST) Facility for Support of Advanced Reactor Technologies

    SciTech Connect

    James E. O'Brien; Piyush Sabharwall; SuJong Yoon

    2001-11-01

    Effective and robust high temperature heat transfer systems are fundamental to the successful deployment of advanced reactors for both power generation and non-electric applications. Plant designs often include an intermediate heat transfer loop (IHTL) with heat exchangers at either end to deliver thermal energy to the application while providing isolation of the primary reactor system. In order to address technical feasibility concerns and challenges a new high-temperature multi-fluid, multi-loop test facility “Advanced Reactor Technology Integral System Test facility” (ARTIST) is under development at the Idaho National Laboratory. The facility will include three flow loops: high-temperature helium, molten salt, and steam/water. Details of some of the design aspects and challenges of this facility, which is currently in the conceptual design phase, are discussed

  17. Vertical Pretreatment Reactor System Two-vessel system for primary and secondary pretreatment at different temperatures

    E-print Network

    Vertical Pretreatment Reactor System Two-vessel system for primary and secondary pretreatment in the pressurized mixing tube · Preheated, premixed biomass is retained for specified residence time in vertical · Residence time is adjusted by changing amount of material held in vertical vessel relative to continuous

  18. FAFTRCS: an experiment in computerized reactor safety systems

    SciTech Connect

    Chisholm, G.H.

    1985-01-01

    Nuclear Power Plant availability and reliability could be improved by the integration of computers into the control environment. However, computer-based systems are historically viewed as being unreliable. This places a burden upon the designer to demonstrate adequate reliability and availability for the computer. The complexity associated with computers coupled with the manual nature of these demonstrations results in a high cost which typically has been justified for critical applications only. This paper investigates a methodology for automating this process and discusses a project which intends to apply this methodology to design verification and validation for a control system which will be installed and tested in an actual reactor control environment. 7 refs., 4 figs., 1 tab.

  19. A RESEARCH ON SEAMLESS PLATFORM CHANGE OF REACTOR PROTECTION SYSTEM FROM PLC TO FPGA

    E-print Network

    A RESEARCH ON SEAMLESS PLATFORM CHANGE OF REACTOR PROTECTION SYSTEM FROM PLC TO FPGA JUNBEOM YOO1-critical systems such as RPS (Reactor Protection System) for nuclear power plants. The software loaded into a PLC development to produce the Verilog program, which is a starting point of typical FPGA devel- opments

  20. A Spouted Bed Reactor Monitoring System for Particulate Nuclear Fuel

    Microsoft Academic Search

    D. S. Wendt; R. L. Bewley; W. E. Windes

    2007-01-01

    Conversion and coating of particle nuclear fuel is performed in spouted (fluidized) bed reactors. The reactor must be capable of operating at temperatures up to 2000°C in inert, flammable, and coating gas environments. The spouted bed reactor geometry is defined by a graphite retort with a 2.5 inch inside diameter, conical section with a 60° included angle, and a 4

  1. Nuclear safety as applied to space power reactor systems

    SciTech Connect

    Cummings, G.E.

    1987-01-01

    Current space nuclear power reactor safety issues are discussed with respect to the unique characteristics of these reactors. An approach to achieving adequate safety and a perception of safety is outlined. This approach calls for a carefully conceived safety program which makes uses of lessons learned from previous terrestrial power reactor development programs. This approach includes use of risk analyses, passive safety design features, and analyses/experiments to understand and control off-design conditions. The point is made that some recent accidents concerning terrestrial power reactors do not imply that space power reactors cannot be operated safety.

  2. Ultrasonic processing of dairy systems in large scale reactors.

    PubMed

    Zisu, Bogdan; Bhaskaracharya, Raman; Kentish, Sandra; Ashokkumar, Muthupandian

    2010-08-01

    High intensity low frequency ultrasound was used to process dairy ingredients to improve functional properties. Based on a number of lab-scale experiments, several experimental parameters were optimised for processing large volumes of whey and casein-based dairy systems in pilot scale ultrasonic reactors. A continuous sonication process at 20 kHz capable of delivering up to 4 kW of power with a flow-through reactor design was used to treat dairy ingredients at flow rates ranging from 200 to 6000 mL/min. Dairy ingredients treated by ultrasound included reconstituted whey protein concentrate (WPC), whey protein and milk protein retentates and calcium caseinate. The sonication of solutions with a contact time of less than 1 min and up to 2.4 min led to a significant reduction in the viscosity of materials containing 18% to 54% (w/w) solids. The viscosity of aqueous dairy ingredients treated with ultrasound was reduced by between 6% and 50% depending greatly on the composition, processing history, acoustic power and contact time. A notable improvement in the gel strength of sonicated and heat coagulated dairy systems was also observed. When sonication was combined with a pre-heat treatment of 80 degrees C for 1 min or 85 degrees C for 30s, the heat stability of the dairy ingredients containing whey proteins was significantly improved. The effect of sonication was attributed mainly to physical forces generated through acoustic cavitation as supported by particle size reduction in response to sonication. As a result, the gelling properties and heat stability aspects of sonicated dairy ingredients were maintained after spray drying and reconstitution. Overall, the sonication procedure for processing dairy systems may be used to improve process efficiency, improve throughput and develop value added ingredients with the potential to deliver economical benefits to the dairy industry. PMID:19948420

  3. Fault tree analysis of the EBR-II reactor shutdown system

    Microsoft Academic Search

    S. A. Kamal; D. J. Hill

    1992-01-01

    As part of the level I Probabilistic Risk Assessment of the Experimental Breeder Reactor II (EBR-II), detailed fault trees for the reactor shutdown system are developed. Fault tree analysis is performed for two classes of transient events that are of particular importance to EBR-II operation: loss-of-flow and transient-overpower. In all parts of EBR-II reactor shutdown system, redundancy has been utilized

  4. Flow system for fish freshness determination based on double multi-enzyme reactor electrodes

    Microsoft Academic Search

    Hirokazu Okuma; Etsuo Watanabe

    2002-01-01

    A double reactor system for the determination of fish and shellfish freshness using the freshness indicator, K-value (K={(HxR+Hx)\\/(ATP+ADP+AMP+IMP+HxR+Hx)}×100), was developed, where ATP, ADP, AMP, IMP, HxR and Hx are adenosine triphosphate, adenosine diphosphate, adenosine monophosphate, inosine monophosphate, inosine and hypoxanthine, respectively. The system consisted of a pair of enzyme reactors with an oxygen electrode positioned close to the respective reactor.

  5. Lunar Regolith Simulant Feed System for a Hydrogen Reduction Reactor System

    NASA Technical Reports Server (NTRS)

    Mueller, R. P.; Townsend, Ivan I., III

    2009-01-01

    One of the goals of In-Situ Resource Utilization (ISRU) on the moon is to produce oxygen from the lunar regolith which is present in the form of Ilmenite (FeTi03) and other compounds. A reliable and attainable method of extracting some of the oxygen from the lunar regolith is to use the hydrogen reduction process in a hot reactor to create water vapor which is then condensed and electrolyzed to obtain oxygen for use as a consumable. One challenge for a production system is to reliably acquire the regolith with an excavator hauler mobility platform and then introduce it into the reactor inlet tube which is raised from the surface and above the reactor itself. After the reaction, the hot regolith (-1000 C) must be expelled from the reactor for disposal by the excavator hauler mobility system. In addition, the reactor regolith inlet and outlet tubes must be sealed by valves during the reaction in order to allow collection of the water vapor by the chemical processing sub-system. These valves must be able to handle abrasive regolith passing through them as well as the heat conduction from the hot reactor. In 2008, NASA has designed and field tested a hydrogen reduction system called ROxygen in order to demonstrate the feasibility of extracting oxygen from lunar regolith. The field test was performed with volcanic ash known as Tephra on Mauna Kea volcano on the Big Island of Hawai'i. The tephra has similar properties to lunar regolith, so that it is regarded as a good simulant for the hydrogen reduction process. This paper will discuss the design, fabrication, operation, test results and lessons learned with the ROxygen regolith feed system as tested on Mauna Kea in November 2008.

  6. Testing of an advanced thermochemical conversion reactor system

    SciTech Connect

    Not Available

    1990-01-01

    This report presents the results of work conducted by MTCI to verify and confirm experimentally the ability of the MTCI gasification process to effectively generate a high-quality, medium-Btu gas from a wider variety of feedstock and waste than that attainable in air-blown, direct gasification systems. The system's overall simplicity, due to the compact nature of the pulse combustor, and the high heat transfer rates attainable within the pulsating flow resonance tubes, provide a decided and near-term potential economic advantage for the MTCI indirect gasification system. The primary objective of this project was the design, construction, and testing of a Process Design Verification System for an indirectly heated, thermochemical fluid-bed reactor and a pulse combustor an an integrated system that can process alternative renewable sources of energy such as biomass, black liquor, municipal solid waste and waste hydrocarbons, including heavy oils into a useful product gas. The test objectives for the biomass portion of this program were to establish definitive performance data on biomass feedstocks covering a wide range of feedstock qualities and characteristics. The test objectives for the black liquor portion of this program were to verify the operation of the indirect gasifier on commercial black liquor containing 65 percent solids at several temperature levels and to characterize the bed carbon content, bed solids particle size and sulfur distribution as a function of gasification conditions. 6 refs., 59 figs., 29 tabs.

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

    NASA Technical Reports Server (NTRS)

    Sinclair, J. H.

    1974-01-01

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

  8. IAEA coordinated research activities on materials for advanced reactor systems

    NASA Astrophysics Data System (ADS)

    Zeman, A.; Inozemtsev, V.; Kamendje, R.; Beatty, R. L.

    2013-11-01

    After the recent accident at the Fukushima Daiichi Nuclear Power Plant, public resentment towards nuclear energy is very high; however it is also important to emphasise that for other facilities the safety record has been remarkably good when compared to those of other new or conventional energy technologies. In addition to clear safety improvements new systems will have increased thermal efficiency, maximised fuel use, and reduced nuclear waste production. In order to initiate commercial deployment of power reactors, small scale demonstrations of such new systems are urgently needed. This will help to develop, test and qualify new structural materials with improved properties with respect to radiation, corrosion, thermal and other degradation processes. To solve all challenges related to the performance parameters of such materials, internationally driven efforts must focus on research, targeted testing, and final selection of appropriate materials. This is recognised as a key milestone in successful demonstration and future deployment of newly designed nuclear reactors. Because of clear synergies between fusion and fission research and development communities have been identified, closer cooperation of research groups has been stimulated. Although some operational conditions are expected to change, many basic features will remain similar. In addition to the material science effort, new experimental facilities are being developed for the study of high-radiation damage effects on the microstructure of candidate materials prior to their qualification. During last 5 years, the International Atomic Energy Agency (IAEA) launched several coordinated research activities in this specific, but very important field. This paper gives a summary of on-going IAEA activities related to the development and characterisation of structural and plasma facing materials for nuclear energy.

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

    SciTech Connect

    Karpov, A. S. [St Petersburg State Polytechnical University, JSC 'System Operator of the United Power System', Leningradskoe RDU (Russian Federation)] [St Petersburg State Polytechnical University, JSC 'System Operator of the United Power System', Leningradskoe RDU (Russian Federation)

    2013-01-15

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

  10. ITM Syngas and ITM H2: Engineering Development of Ceramic Membrane Reactor Systems for

    E-print Network

    ITM Syngas and ITM H2: Engineering Development of Ceramic Membrane Reactor Systems for Converting, refineries. Purified hydrogen can be liquefied and transported to the point of use and vaporized is to research, develop and demonstrate a novel ceramic membrane reactor system for the low-cost conversion

  11. SP-100 space reactor power system readiness and mission flexibility

    NASA Astrophysics Data System (ADS)

    Josloff, Allan T.; Matteo, Donald N.; Bailey, H. Sterling

    1993-01-01

    The SP-100 Space Reactor Power System (SRPS) is being developed by GE, under contract to the U.S. Department of Energy, to provide electrical power in the range of 10s to 100s of kW. The system represents an enabling technology for a wide variety of earth orbital and interplanetary science missions, nuclear electric propulsion (NEP) stages, and lunar/Mars surface power for the Space Exploration Initiative (SEI). An effective infracture of Industry, National Laboratories and Government agencies has made substantial progress since the 1988 System Design Review. Hardware development and testing has progressed to the point of resolving the key technical feasibility issues. The technology and design is now at a state of readiness to support the definition of early flight demonstration missions. The benefits of utilizing a low power (6 to 20 kWe range) early flight mission as a precursor to operational missions in the 100 kWe range has received renewed interest among Government Agencies and Industry. Studies and assessments were performed throughout 1992 to further refine the potential missions and the SP-100 Space Reactor Power Systems that could be available to support these missions. The results of assessment showed that the ``first generation'' technology available now from the SP-100 program can support a wide range of candidate missions. The status of the nuclear technology was matured to the level of supporting a flight design with the present available data base. The conductively coupled thermoelectric cell technology is now in the cell level testing and verification phase and component level readiness is projected to be complete by the end of GFY94. Power system designs using the present day flight proven RTG unicouple have been established and also represent an attractive option for early launches. These design concepts are discussed in further detail in a companion paper. (Josloff 1993). This paper will review the SP-100 key features, technology status and early flight mission readiness and updates an earlier paper on this topic (Josloff 1992a).

  12. Secondary Startup Neutron Sources as a Source of Tritium in a Pressurized Water Reactor (PWR) Reactor Coolant System (RCS)

    SciTech Connect

    Shaver, Mark W.; Lanning, Donald D.

    2010-02-01

    The hypothesis of this paper is that the Zircaloy clad fuel source is minimal and that secondary startup neutron sources are the significant contributors of the tritium in the RCS that was previously assigned to release from fuel. Currently there are large uncertainties in the attribution of tritium in a Pressurized Water Reactor (PWR) Reactor Coolant System (RCS). The measured amount of tritium in the coolant cannot be separated out empirically into its individual sources. Therefore, to quantify individual contributors, all sources of tritium in the RCS of a PWR must be understood theoretically and verified by the sum of the individual components equaling the measured values.

  13. System simulation of a multicell thermionic space power reactor

    NASA Astrophysics Data System (ADS)

    von Arx, Alan Vincent

    For many years, thermionic power has been considered for space application. The prominent feature of the power conversion system is that there are no moving parts. Although designs have been developed by various organizations, no comprehensive system models are known to exist which can simulate transient behavior of a multicell design nor is there a method to directly couple these models to other codes that can calculate variations in reactivity. Thus, a procedure has been developed to couple the performance calculations of a space nuclear reactor thermal/hydraulics code with a neutron diffusion code to analyze temperature feedback. Thermionic power is based on the thermionic emissions principle where free electrons in a conductor have sufficient energy to escape the surface. Kinetic energy is given to the electrons by heating the conductor. Specifically, a 48 kWe thermionic power converter system model has been developed and used to model startup and other transients. Less than 10% of the fuel heat is converted to electricity, and the rest is rejected to space via a heat pipe radiator. An electromagnetic pump circulates the liquid metal coolant. First, a startup transient model was developed which showed stable operation through ignition of the Thermionic Fuel Elements (TFEs) and thawing of the radiator heat pipes. Also, the model's capability was expanded to include two-phase heat transfer to model boiling using coupled mass and thermal energy conservation equations. The next step incorporated effects of reactivity feedback---showing that various mechanisms will prevent power and temperature run-up for a flow reduction scenario where the reactor control systems fail to respond. In particular, the Doppler effect was shown to counter a positive worth due to partial core voiding although steps must be taken to preclude film boiling in that high superheats will result in TFE failures. Finally, analysis of the core grid spacer location suggests it should be located at the core outlet only. Applicable operational data were also assessed for TOPAZ II tests. A steady-state analysis showed a good comparison with other modeling codes, and TFE performance agreed within 3% of the experimental data---thus, validating the performance calculations.

  14. Characterization of bacterial communities in hybrid upflow anaerobic sludge blanket (UASB)-membrane bioreactor (MBR) process for berberine antibiotic wastewater treatment.

    PubMed

    Qiu, Guanglei; Song, Yong-Hui; Zeng, Ping; Duan, Liang; Xiao, Shuhu

    2013-08-01

    Biodegradation of berberine antibiotic was investigated in upflow anaerobic sludge blanket (UASB)-membrane bioreactor (MBR) process. After 118days of operation, 99.0%, 98.0% and 98.0% overall removals of berberine, COD and NH4(+)-N were achieved, respectively. The detailed composition of the established bacterial communities was studied by using 16S rDNA clone library. Totally, 400 clones were retrieved and grouped into 186 operational taxonomic units (OTUs). UASB was dominated by Firmicutes and Bacteroidetes, while Proteobacteria, especially Alpha- and Beta-proteobacteria were prevalent in the MBRs. Clostridium, Eubacterium and Synergistes in the UASB, as well as Hydrogenophaga, Azoarcus, Sphingomonas, Stenotrophomonas, Shinella and Alcaligenes in the MBRs were identified as potential functional species in biodegradation of berberine and/or its metabolites. The bacterial community compositions in two MBRs were significantly discrepant. However, the identical functions of the functional species ensured the comparable pollutant removal performances in two bioreactors. PMID:23735790

  15. Incorporating ''fuzzy'' data and logical relations in the design of expert systems for nuclear reactors

    SciTech Connect

    Guth, M.A.S.

    1987-01-01

    This paper applies the method of assigning probability in Dempster-Shafer Theory (DST) to the components of rule-based expert systems used in the control of nuclear reactors. Probabilities are assigned to premises, consequences, and rules themselves. This paper considers how uncertainty can propagate through a system of Boolean equations, such as fault trees or expert systems. The probability masses assigned to primary initiating events in the expert system can be derived from observing a nuclear reactor in operation or based on engineering knowledge of the reactor parts. Use of DST mass assignments offers greater flexibility to the construction of expert systems.

  16. A long term radiological risk model for plutonium-fueled and fission reactor space nuclear system

    SciTech Connect

    Bartram, B.W.; Dougherty, D.K.

    1987-01-01

    This report describes the optimization of the RISK III mathematical model, which provides risk assessment for the use of a plutonium-fueled, fission reactor in space systems. The report discusses possible scenarios leading to radiation releases on the ground; distinctions are made for an intact reactor and a dispersed reactor. Also included are projected dose equivalents for various accident situations. 54 refs., 31 figs., 11 tabs. (TEM)

  17. Symbiotic system of a fusion and a fission reactor with very simple fuel reprocessing

    Microsoft Academic Search

    V. L. Blinkin; V. M. Novikov

    1978-01-01

    The paper discusses a symbiotic fusion and fission reactor system. The method is based on producing U-233 in the blanket of the fusion reactor from thorium which circulates as ThF4 in a mixture of sodium and beryllium fluoride melts. The U-233 produced in the blanket supplies a fission reactor and generates fuel for additional symbiotic installations. Electric power and breed

  18. OUTLINE OF CHEMICAL ENGINEERING 4K3/6K3: Sep-Dec, 2013 Reactor Design for Heterogenous Systems

    E-print Network

    Thompson, Michael

    an understanding of Advanced Reactor Design including Catalytic kinetics, mass transfer limitations, packed and fluidized bed reactors and two phase reactors. POLICY REMINDER: Academic Integrity: You are expectedOUTLINE OF CHEMICAL ENGINEERING 4K3/6K3: Sep-Dec, 2013 Reactor Design for Heterogenous Systems

  19. Experimental Studies of NGNP Reactor Cavity Cooling System With Water

    SciTech Connect

    Corradini, Michael; Anderson, Mark; Hassan, Yassin; Tokuhiro, Akira

    2013-01-16

    This project will investigate the flow behavior that can occur in the reactor cavity cooling system (RCCS) with water coolant under the passive cooling-mode of operation. The team will conduct separate-effects tests and develop associated scaling analyses, and provide system-level phenomenological and computational models that describe key flow phenomena during RCCS operation, from forced to natural circulation, single-phase flow and two-phase flow and flashing. The project consists of the following tasks: Task 1. Conduct separate-effects, single-phase flow experiments and develop scaling analyses for comparison to system-level computational modeling for the RCCS standpipe design. A transition from forced to natural convection cooling occurs in the standpipe under accident conditions. These tests will measure global flow behavior and local flow velocities, as well as develop instrumentation for use in larger scale tests, thereby providing proper flow distribution among standpipes for decay heat removal. Task 2. Conduct separate-effects experiments for the RCCS standpipe design as two-phase flashing occurs and flow develops. As natural circulation cooling continues without an ultimate heat sink, water within the system will heat to temperatures approaching saturation , at which point two-phase flashing and flow will begin. The focus is to develop a phenomenological model from these tests that will describe the flashing and flow stability phenomena. In addition, one could determine the efficiency of phase separation in the RCCS storage tank as the two-phase flashing phenomena ensues and the storage tank vents the steam produced. Task 3. Develop a system-level computational model that will describe the overall RCCS behavior as it transitions from forced flow to natural circulation and eventual two-phase flow in the passive cooling-mode of operation. This modeling can then be used to test the phenomenological models developed as a function of scale.

  20. Performance of UASB septic tank for treatment of concentrated black water within DESAR concept.

    PubMed

    Kujawa-Roeleveld, K; Fernandes, T; Wiryawan, Y; Tawfik, A; Visser, M; Zeeman, G

    2005-01-01

    Separation of wastewater streams produced in households according to their origin, degree of pollution and affinity to a specific treatment constitutes a starting point in the DESAR concept (decentralised sanitation and reuse). Concentrated black water and kitchen waste carry the highest load of organic matter and nutrients from all waste(water)streams generated from different human activities. Anaerobic digestion of concentrated black water is a core technology in the DESAR concept. The applicability of the UASB septic tank for treatment of concentrated black water was investigated under two different temperatures, 15 and 25 degrees C. The removal of total COD was dependent on the operational temperature and attained 61 and 74% respectively. A high removal of the suspended COD of 88 and 94% respectively was measured. Effluent nutrients were mainly in the soluble form. Precipitation of phosphate was observed. Effective sludge/water separation, long HRT and higher operational temperature contributed to a reduction of E. coli. Based on standards there is little risk of contamination with heavy metals when treated effluent is to be applied in agriculture as fertiliser. PMID:16180443

  1. On-line test of power distribution prediction system for boiling water reactors

    Microsoft Academic Search

    Y. Nishizawa; T. Kiguchi; S. Kobayashi; K. Takumi; H. Tanaka; R. Tsutsumi; M. Yokomi

    1982-01-01

    A power distribution prediction system for boiling water reactors has been developed and its on-line performance test has proceeded at an operating commercial reactor. This system predicts the power distribution or thermal margin in advance of control rod operations and core flow rate change. This system consists of an on-line computer system, an operator's console with a color cathode-ray tube,

  2. Passive cooling system for top entry liquid metal cooled nuclear reactors

    DOEpatents

    Boardman, Charles E. (Saratoga, CA); Hunsbedt, Anstein (Los Gatos, CA); Hui, Marvin M. (Cupertino, CA)

    1992-01-01

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

  3. Shielding considerations for advanced space nuclear reactor systems

    SciTech Connect

    Angelo, J.P. Jr.; Buden, D.

    1982-01-01

    To meet the anticipated future space power needs, the Los Alamos National Laboratory is developing components for a compact, 100 kW/sub e/-class heat pipe nuclear reactor. The reactor uses uranium dioxide (UO/sub 2/) as its fuel, and is designed to operate around 1500 k. Heat pipes are used to remove thermal energy from the core without the use of pumps or compressors. The reactor heat pipes transfer mal energy to thermoelectric conversion elements that are advanced versions of the converters used on the enormously successful Voyager missions to the outer planets. Advanced versions of this heat pipe reactor could also be used to provide megawatt-level power plants. The paper reviews the status of this advanced heat pipe reactor and explores the radiation environments and shielding requirements for representative manned and unmanned applications.

  4. Hanging core support system for a nuclear reactor. [LMFBR

    DOEpatents

    Burelbach, J.P.; Kann, W.J.; Pan, Y.C.; Saiveau, J.G.; Seidensticker, R.W.

    1984-04-26

    For holding the reactor core in the confining reactor vessel, a support is disclosed that is structurally independent of the vessel, that is dimensionally accurate and stable, and that comprises tandem tension linkages that act redundantly of one another to maintain stabilized core support even in the unlikely event of the complete failure of one of the linkages. The core support has a mounting platform for the reactor core, and unitary structure including a flange overlying the top edge of the reactor vessels, and a skirt and box beams between the flange and platform for establishing one of the linkages. A plurality of tension rods connect between the deck closing the reactor vessel and the platform for establishing the redundant linkage. Loaded Belleville springs flexibly hold the tension rods at the deck and separable bayonet-type connections hold the tension rods at the platform.

  5. Analysis of a microbial community associated with polychlorinated biphenyl degradation in anaerobic batch reactors.

    PubMed

    Gomes, B C; Adorno, M A T; Okada, D Y; Delforno, T P; Lima Gomes, P C F; Sakamoto, I K; Varesche, M B A

    2014-11-01

    The degradation of polychlorinated biphenyls (PCBs) was investigated under fermentative-methanogenic conditions for up to 60 days in the presence of anaerobic biomass from a full-scale UASB reactor. The low methane yields in the PCBs-spiked batch reactors suggested that the biomass had an inhibitory effect on the methanogenic community. Reactors containing PCBs and co-substrates (ethanol/sodium formate) exhibited substantial PCB reductions from 0.7 to 0.2 mg mL(-1). For the Bacteria domain, the PCBs-spiked reactors were grouped with the PCB-free reactors with a similarity of 55 %, which suggested the selection of a specific population in the presence of PCBs. Three genera of bacteria were found exclusively in the PCB-spiked reactors and were identified using pyrosequencing analysis, Sedimentibacter, Tissierela and Fusibacter. Interestingly, the Sedimentibacter, which was previously correlated with the reductive dechlorination of PCBs, had the highest relative abundance in the RCS-PCB (7.4 %) and RCS-PCB-PF (12.4 %) reactors. Thus, the anaerobic sludge from the UASB reactor contains bacteria from the Firmicutes phylum that are capable of degrading PCBs. PMID:25104219

  6. Reactor physic and reprocessing scheme for innovative molten salt reactor system

    Microsoft Academic Search

    S. Delpech; E. Merle-Lucotte; D. Heuer; M. Allibert; V. Ghetta; C. Le-Brun; X. Doligez; G. Picard

    2009-01-01

    The molten salt reactor is one of the six concepts retained by the Generation IV forum in 2001. Based on the MSRE and MSBR concepts developed by ORNL in the 60s which involve a liquid fuel constituted of fluorine molten salt at a temperature close to 600°C, new developments with innovative approach and technology have been realized which contribute to

  7. Reactor test: testing system for periodical inspection of reactor vessels for nuclear plant

    Microsoft Academic Search

    Junghem

    1972-01-01

    The demand for periodical inspection of reactor vessels for nuaclear ; power plants has caused a partially new inspection technology to be developed. ; The inspection is mainly made by means of mechanized, remote controlled ; inspection devices and by using nondestructive testing, mainly by ultrasonics, ; and visual inspection by television camera. The purpose for the inspection is to

  8. 78 FR 64029 - Cost-Benefit Analysis for Radwaste Systems for Light-Water-Cooled Nuclear Power Reactors

    Federal Register 2010, 2011, 2012, 2013, 2014

    2013-10-25

    ...Cost-Benefit Analysis for Radwaste Systems for Light-Water-Cooled Nuclear Power Reactors...Cost-Benefit Analysis for Radwaste Systems for Light-Water-Cooled Nuclear Power Reactors...gaseous radwaste system components for light water nuclear power reactors....

  9. Quantification of anammox activity in a denitrification reactor for a recirculating aquaculture system

    Microsoft Academic Search

    Ori Lahav; Iris Bar Massada; Dimitry Yackoubov; Ruth Zelikson; Noam Mozes; Yossi Tal; Sheldon Tarre

    2009-01-01

    The activity of anammox bacteria in a denitrification reactor in a recirculating aquaculture system (RAS) for gilthead seabream production was investigated. Organic matter, extracted from the pond's solid filter, was used as the electron donor and carbon source for the denitrification reaction. The reactor was operated at four solid retention times (SRT). At steady state, anammox activity showed similar activity

  10. APPLICATIONS ANALYSIS REPORT: ECO LOGIC INTERNATIONAL GAS-PHASE CHEMICAL REDUCTION PROCESS - THE REACTOR SYSTEM

    EPA Science Inventory

    This report details the Superfund Innovative Technology Evaluation of Eco Logic International's gas-phase chemical reduction process, with an emphasis on their Reactor System. he Eco Logic process employees a high temperature reactor filled with hydrogen gas as the means to destr...

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

    NASA Technical Reports Server (NTRS)

    Lantz, E.; Mayo, W.

    1972-01-01

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

  12. Hybrid energy systems (HESs) using small modular reactors (SMRs)

    SciTech Connect

    S. Bragg-Sitton

    2014-10-01

    Large-scale nuclear reactors are traditionally operated for a singular purpose: steady-state production of dispatchable baseload electricity that is distributed broadly on the electric grid. While this implementation is key to a sustainable, reliable energy grid, small modular reactors (SMRs) offer new opportunities for increased use of clean nuclear energy for both electric and thermal ap plications in more locations – while still accommodating the desire to support renewable production sources.

  13. Sliding mode control of the space nuclear reactor system TOPAZ II

    SciTech Connect

    Shtessel, Y.B. [Department of Electrical and Computer Engineering, University of Alabama in Huntsville, Huntsville, Alabama 35899 (United States); Wyant, F.J. [Phillips Laboratory/Power Management Division, 3550 Aberdeen Avenue, SE, Albuquerque, New Mexico, 87117-5776 (United States)

    1996-03-01

    The Automatic Control System (ACS) of the space nuclear reactor power system TOPAZ II that generates electricity from nuclear heat using in-core thermionic converters is considered. Sliding Mode Control Technique was applied to the reactor system controller design in order to provide the robust high accuracy following of a neutron (thermal) power reference profile in a start up regime and a payload electric power (current) reference profile following in an operation regime. Extensive simulations of the TOPAZ II reactor system with the designed sliding mode controllers showed improved accuracy and robustness of the reactor system performances in a start up regime and in an electric power supply regime as well. {copyright} {ital 1996 American Institute of Physics.}

  14. Neutron Density Fluctuations in Point Reactor Systems with Dichotomic Reactivity Noise

    Microsoft Academic Search

    Okitsugu SAKO

    1984-01-01

    The exactly solvable stochastic point reactor model systems are analyzed through the stochastic Liouville equation. Three kinds of model systems are treated: (1) linear system without delayed neutrons, (2) linear system with one-group of delayed neutrons, and (3) nonlinear system with direct power feedback. The exact expressions for the fluctuations of neutron density, such as the moments, autocorrelation function and

  15. Analyzing TRV of CB When Installing Current Limit Reactors in UHV Power Systems

    Microsoft Academic Search

    H. S. Park; J. W. Woo; J. W. Kang; K. S. Han; S. O. Han

    Due to the high-capacity and complexity in power systems, an increase of high transmission capacity lowers the system impedance of the power system. Generally, according to the lower equivalent impedance of the system, system stability will be improved, but fault current also increases. Regarding high-fault currents, KEPCO (Korea Electric Power Corperation) plans to use CLR (Current Limit Reactors) on UHV

  16. Physical modelling of the composting environment: A review. Part 1: Reactor systems

    SciTech Connect

    Mason, I.G. [Department of Civil Engineering, University of Canterbury, Private Bag 4800, Christchurch (New Zealand)]. E-mail: ian.mason@canterbury.ac.nz; Milke, M.W. [Department of Civil Engineering, University of Canterbury, Private Bag 4800, Christchurch (New Zealand)

    2005-07-01

    In this paper, laboratory- and pilot-scale reactors used for investigation of the composting process are described and their characteristics and application reviewed. Reactor types were categorised by the present authors as fixed-temperature, self-heating, controlled temperature difference and controlled heat flux, depending upon the means of management of heat flux through vessel walls. The review indicated that fixed-temperature reactors have significant applications in studying reaction rates and other phenomena, but may self-heat to higher temperatures during the process. Self-heating laboratory-scale reactors, although inexpensive and uncomplicated, were shown to typically suffer from disproportionately large losses through the walls, even with substantial insulation present. At pilot scale, however, even moderately insulated self-heating reactors are able to reproduce wall losses similar to those reported for full-scale systems, and a simple technique for estimation of insulation requirements for self-heating reactors is presented. In contrast, controlled temperature difference and controlled heat flux laboratory reactors can provide spatial temperature differentials similar to those in full-scale systems, and can simulate full-scale wall losses. Surface area to volume ratios, a significant factor in terms of heat loss through vessel walls, were estimated by the present authors at 5.0-88.0 m{sup 2}/m{sup 3} for experimental composting reactors and 0.4-3.8 m{sup 2}/m{sup 3} for full-scale systems. Non-thermodynamic factors such as compression, sidewall airflow effects, channelling and mixing may affect simulation performance and are discussed. Further work to investigate wall effects in composting reactors, to obtain more data on horizontal temperature profiles and rates of biological heat production, to incorporate compressive effects into experimental reactors and to investigate experimental systems employing natural ventilation is suggested.

  17. Treatment of fruit-juice industry wastewater in a two-stage anaerobic hybrid (AH) reactor system followed by a sequencing batch reactor (SBR)

    Microsoft Academic Search

    A. Tawfik; H. El-Kamah

    2012-01-01

    This study has been carried out to assess the performance of a combined system consisting of an anaerobic hybrid (AH) reactor followed by a sequencing batch reactor (SBR) for treatment of fruit-juice industry wastewater at a temperature of 26 °C. Three experimental runs were conducted in this investigation. In the first experiment, a single-stage AH reactor was operated at a hydraulic

  18. Treatment of fruit-juice industry wastewater in a two-stage anaerobic hybrid (AH) reactor system followed by a sequencing batch reactor (SBR)

    Microsoft Academic Search

    A. Tawfik; H. El-Kamah

    2011-01-01

    This study has been carried out to assess the performance of a combined system consisting of an anaerobic hybrid (AH) reactor followed by a sequencing batch reactor (SBR) for treatment of fruit-juice industry wastewater at a temperature of 26 °C. Three experimental runs were conducted in this investigation. In the first experiment, a single-stage AH reactor was operated at a hydraulic

  19. Advanced High-Temperature Reactor Dynamic System Model Development: April 2012 Status

    SciTech Connect

    Qualls, A.L.; Cetiner, M.S.; Wilson, T.L., Jr.

    2012-04-30

    The Advanced High-Temperature Reactor (AHTR) is a large-output fluoride-salt-cooled high-temperature reactor (FHR). An early-phase preconceptual design of a 1500 MW(e) power plant was developed in 2011 [Refs. 1 and 2]. An updated version of this plant is shown as Fig. 1. FHRs feature low-pressure liquid fluoride salt cooling, coated-particle fuel, a high-temperature power cycle, and fully passive decay heat rejection. The AHTR is designed to be a “walk away” reactor that requires no action to prevent large off-site releases following even severe reactor accidents. This report describes the development of dynamic system models used to further the AHTR design toward that goal. These models predict system response during warmup, startup, normal operation, and limited off-normal operating conditions. Severe accidents that include a loss-of-fluid inventory are not currently modeled. The scope of the models is limited to the plant power system, including the reactor, the primary and intermediate heat transport systems, the power conversion system, and safety-related or auxiliary heat removal systems. The primary coolant system, the intermediate heat transport system and the reactor building structure surrounding them are shown in Fig. 2. These systems are modeled in the most detail because the passive interaction of the primary system with the surrounding structure and heat removal systems, and ultimately the environment, protects the reactor fuel and the vessel from damage during severe reactor transients. The reactor silo also plays an important role during system warmup. The dynamic system modeling tools predict system performance and response. The goal is to accurately predict temperatures and pressures within the primary, intermediate, and power conversion systems and to study the impacts of design changes on those responses. The models are design tools and are not intended to be used in reactor qualification. The important details to capture in the primary system relate to flows within the reactor vessel during severe events and the resulting temperature profiles (temperature and duration) for major components. Critical components include the fuel, reactor vessel, primary piping, and the primary-to-intermediate heat exchangers (P-IHXs). The major AHTR power system loops are shown in Fig. 3. The intermediate heat transfer system is a group of three pumped salt loops that transports the energy produced in the primary system to the power conversion system. Two dynamic system models are used to analyze the AHTR. A Matlab/Simulink?-based model initiated in 2011 has been updated to reflect the evolving design parameters related to the heat flows associated with the reactor vessel. The Matlab model utilizes simplified flow assumptions within the vessel and incorporates an empirical representation of the Direct Reactor Auxiliary Cooling System (DRACS). A Dymola/Modelica? model incorporates a more sophisticated representation of primary coolant flow and a physics-based representation of the three-loop DRACS thermal hydraulics. This model is not currently operating in a fully integrated mode. The Matlab model serves as a prototype and provides verification for the Dymola model, and its use will be phased out as the Dymola model nears completion. The heat exchangers in the system are sized using spreadsheet-based, steady-state calculations. The detail features of the heat exchangers are programmed into the dynamic models, and the overall dimensions are used to generate realistic plant designs. For the modeling cases where the emphasis is on understanding responses within the intermediate and primary systems, the power conversion system may be modeled as a simple boundary condition at the intermediate-to-power conversion system heat exchangers.

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

    NASA Technical Reports Server (NTRS)

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

    1974-01-01

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

  1. The development of a remote monitoring system for the Nuclear Science Center reactor

    E-print Network

    Jiltchenkov, Dmitri Victorovich

    2002-01-01

    With funding provided by Nuclear Energy Research Initiative (NERI), design of Secure, Transportable, Autonomous Reactors (STAR) to aid countries with insufficient energy supplies is underway. The development of a new monitoring system that allows...

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

    NASA Technical Reports Server (NTRS)

    Davis, H. S.

    1984-01-01

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

  3. Passive decay heat removal system for water-cooled nuclear reactors

    DOEpatents

    Forsberg, Charles W. (Oak Ridge, TN)

    1991-01-01

    A passive decay-heat removal system for a water-cooled nuclear reactor employs a closed heat transfer loop having heat-exchanging coils inside an open-topped, insulated box located inside the reactor vessel, below its normal water level, in communication with a condenser located outside of containment and exposed to the atmosphere. The heat transfer loop is located such that the evaporator is in a position where, when the water level drops in the reactor, it will become exposed to steam. Vapor produced in the evaporator passes upward to the condenser above the normal water level. In operation, condensation in the condenser removes heat from the system, and the condensed liquid is returned to the evaporator. The system is disposed such that during normal reactor operations where the water level is at its usual position, very little heat will be removed from the system, but during emergency, low water level conditions, substantial amounts of decay heat will be removed.

  4. Experimental and Computational Study of a Scaled Reactor Cavity Cooling System

    E-print Network

    Vaghetto, Rodolfo

    2013-11-25

    safety system that will be incorporated in the VTHR, designed to remove the heat from the reactor cavity and maintain the temperature of structures and concrete walls under desired limits during normal operation and accident scenarios. A small scale (1...

  5. Mitigation of steam generator tube rupture in a pressurized water reactor with passive safety systems

    DOEpatents

    McDermott, D.J.; Schrader, K.J.; Schulz, T.L.

    1994-05-03

    The effects of steam generator tube ruptures in a pressurized water reactor are mitigated by reducing the pressure in the primary loop by diverting reactor coolant through the heat exchanger of a passive heat removal system immersed in the in containment refueling water storage tank in response to a high feed water level in the steam generator. Reactor coolant inventory is maintained by also in response to high steam generator level introducing coolant into the primary loop from core make-up tanks at the pressure in the reactor coolant system pressurizer. The high steam generator level is also used to isolate the start-up feed water system and the chemical and volume control system to prevent flooding into the steam header. 2 figures.

  6. Power conversion system design for supercritical carbon dioxide cooled indirect cycle nuclear reactors

    E-print Network

    Gibbs, Jonathan Paul

    2008-01-01

    The supercritical carbon dioxide (S-CO?) cycle is a promising advanced power conversion cycle which couples nicely to many Generation IV nuclear reactors. This work investigates the power conversion system design and ...

  7. Design of Compact and Full-Size Simplified Boiling Water Reactors With Fully Passive Safety Systems

    Microsoft Academic Search

    M. Ishii; S. T. Revankar; Y. Xu

    2002-01-01

    Scientific designs of two next-generation simplified boiling water reactors (SBWRs) namely, a compact modular 200 MWe SBWR and a full-size 1200-MWe SBWR have been developed. The design involved identification of principal design criteria dictated by the safe operation of the reactor, identification of coolant requirements, and the design of the engineered safety and emergency cooling systems based on passive systems.

  8. System pressure effect on the nuclear reactor limiting criterion. Revision 1

    SciTech Connect

    Chen, Kuo-Fu

    1990-12-31

    The acceptable operating limits of a nuclear reactor are set to prevent fuel cladding damage. Critical Heat Flux (CHF) is the limiting criterion for the high pressure systems such as the BWRs (6.9 MPa) and the PWRs (13.8 MPa). However, the Onset of Flow Instability (OFI) is the limiting criterion of the low pressure system such as the existing Savannah River Site (SRS) production reactors (0.2 MPa). The physical basis of this difference is presented. 3 refs.

  9. Probability level of readiness in supervised protective systems for nuclear reactors

    Microsoft Academic Search

    Kontoleon

    1980-01-01

    The readiness of supervised protective systems for nuclear reactors is analyzed by the use of a four-state Markov process. The supervisions are short and reveal the capability of the system to initiate the protection action. When this capability--defined as the probability of initiating the protection action-is found to be below an acceptable level, the reactor is shut down. The analysis

  10. Microbial and Physicochemical Characteristics of Compact Anaerobic Ammonium-Oxidizing Granules in an Upflow Anaerobic Sludge Blanket Reactor ?

    PubMed Central

    Ni, Bing-Jie; Hu, Bao-Lan; Fang, Fang; Xie, Wen-Ming; Kartal, Boran; Liu, Xian-Wei; Sheng, Guo-Ping; Jetten, Mike; Zheng, Ping; Yu, Han-Qing

    2010-01-01

    Anaerobic ammonium oxidation (anammox) is a promising new process to treat high-strength nitrogenous wastewater. Due to the low growth rate of anaerobic ammonium-oxidizing bacteria, efficient biomass retention is essential for reactor operation. Therefore, we studied the settling ability and community composition of the anaerobic ammonium-oxidizing granules, which were cultivated in an upflow anaerobic sludge blanket (UASB) reactor seeded with aerobic granules. With this seed, the start-up period was less than 160 days at a NH4+-N removal efficiency of 94% and a loading rate of 0.064 kg N per kg volatile suspended solids per day. The formed granules were bright red and had a high settling velocity (41 to 79 m h?1). Cells and extracellular polymeric substances were evenly distributed over the anaerobic ammonium-oxidizing granules. The high percentage of anaerobic ammonium-oxidizing bacteria in the granules could be visualized by fluorescent in situ hybridization and electron microscopy. The copy numbers of 16S rRNA genes of anaerobic ammonium-oxidizing bacteria in the granules were determined to be 4.6 × 108 copies ml?1. The results of this study could be used for a better design, shorter start-up time, and more stable operation of anammox systems for the treatment of nitrogen-rich wastewaters. PMID:20190088

  11. Nuclear reactor descriptions for space power systems analysis

    NASA Technical Reports Server (NTRS)

    Mccauley, E. W.; Brown, N. J.

    1972-01-01

    For the small, high performance reactors required for space electric applications, adequate neutronic analysis is of crucial importance, but in terms of computational time consumed, nuclear calculations probably yield the least amount of detail for mission analysis study. It has been found possible, after generation of only a few designs of a reactor family in elaborate thermomechanical and nuclear detail to use simple curve fitting techniques to assure desired neutronic performance while still performing the thermomechanical analysis in explicit detail. The resulting speed-up in computation time permits a broad detailed examination of constraints by the mission analyst.

  12. Tritium Formation and Mitigation in High-Temperature Reactor Systems

    SciTech Connect

    Piyush Sabharwall; Carl Stoots; Hans A. Schmutz

    2013-03-01

    Tritium is a radiologically active isotope of hydrogen. It is formed in nuclear reactors by neutron absorption and ternary fission events and can subsequently escape into the environment. To prevent the tritium contamination of proposed reactor buildings and surrounding sites, this study examines the root causes and potential mitigation strategies for permeation of tritium (such as: materials selection, inert gas sparging, etc...). A model is presented that can be used to predict permeation rates of hydrogen through metallic alloys at temperatures from 450–750 degrees C. Results of the diffusion model are presented for a steady production of tritium

  13. Sodium leak detection system for liquid metal cooled nuclear reactors

    DOEpatents

    Modarres, Dariush (12 La Vista Verde, Rancho Palos Verdes, CA 90274)

    1991-01-01

    A light source is projected across the gap between the containment vessel and the reactor vessel. The reflected light is then analyzed with an absorption spectrometer. The presence of any sodium vapor along the optical path results in a change of the optical transmissivity of the media. Since the absorption spectrum of sodium is well known, the light source is chosen such that the sensor is responsive only to the presence of sodium molecules. The optical sensor is designed to be small and require a minimum of amount of change to the reactor containment vessel.

  14. Designs for remote inspection of the ALMR Reactor Vessel Auxiliary Cooling System (RVACS)

    SciTech Connect

    Sweeney, F.J. (Oak Ridge National Lab., TN (United States)); Carroll, D.G. (General Electric Co., San Jose, CA (United States)); Chen, C. (Tennessee Univ., Knoxville, TN (United States)); Crane, C.; Dalton, R. (Florida Univ., Gainesville, FL (United States)); Taylor, J.R. (Martin Marietta Energy Systems, Inc., Oak Ridge, TN (United States)); Tosunoglu, S. (Texas Univ., Austin, TX (United States))

    1993-01-01

    One of the most important safety systems in General Electric's (GI) Advanced Liquid Metal Reactor (ALMR) is the Reactor Vessel Auxiliary Cooling System (RVACS). Because of high temperature, radiation, and restricted space conditions, GI desired methods to remotely inspect the RVACS, emissive coatings, and reactor vessel welds during normal refueling operations. The DOE/NE Robotics for Advanced Reactors program formed a team to evaluate the ALMR design for remote inspection of the RVACS. Conceptual designs for robots to perform the required inspection tasks were developed by the team. Design criteria for these remote systems included robot deployment, power supply, navigation, environmental hardening of components, tether management, communication with an operator, sensing, and failure recovery. The operation of the remote inspection concepts were tested using 3-D simulation models of the ALMR. In addition, the team performed an extensive technology review of robot components that could survive the environmental conditions in the RVACS.

  15. GE advanced boiling water reactors and plant systems design

    Microsoft Academic Search

    D. R. Wilkins; J. Chang

    1992-01-01

    The nuclear option is becoming increasingly important as the need for new baseload capacity and the environmental impact of burning fossil fuels becomes more evident. Advanced Light Water Reactor nuclear plants which are prelicensed and standardized are key to making nuclear energy an attractive option. GE Nuclear Energy has developed two new advanced and simplified BWR designs aimed at making

  16. Heat insulating system for a fast reactor shield slab

    DOEpatents

    Kotora, J. Jr.; Groh, E.F.; Kann, W.J.; Burelbach, J.P.

    1984-04-10

    Improved thermal insulation for a nuclear reactor deck comprises many helical coil springs disposed in generally parallel, side-by-side laterally overlapping or interfitted relationship to one another so as to define a three-dimensional composite having both metal and voids between the metal, and enclosure means for holding the composite to the underside of the deck.

  17. Heat pipe space nuclear reactor design assessment. Volume 1: Design status of the SP-100 heat pipe space nuclear reactor system

    NASA Astrophysics Data System (ADS)

    Dean, V. F.; El-Genk, M. S.; Louie, D. L. Y.; Woodall, D. M.

    1985-08-01

    This document reviews the design status of the SP-100, heat pipe space nuclear reactor system. It also identifies those systems and components requiring additional research to support continued SP-100 system development. The heat pipe reactor was designed to produce 100 KWe of continuous power in a space environment. The design constraints include an expected system operation time of 7 years and a maximum weight of approx. 3000 kg. The reactor, employing an unclad, highly enriched uranium dioxide core, operates as a fast reactor, and is cooled by high temperature molybdenum -- 13 percent rhenium, heat pipes with lithium working fluid. Electric power is generated by thermoelectric converters, with the bulk of the thermal energy rejected to space by a radiator panel system.

  18. Fossil-fuel processing technical/professional services: comparison of Fischer-Tropsch reactor systems. Phase I, final report

    SciTech Connect

    Thompson, G.J.; Riekena, M.L.; Vickers, A.G.

    1981-09-01

    The Fischer-Tropsch reaction was commercialized in Germany and used to produce military fuels in fixed bed reactors. It was recognized from the start that this reactor system had severe operating and yield limitations and alternative reactor systems were sought. In 1955 the Sasol I complex, using an entrained bed (Synthol) reactor system, was started up in South Africa. Although this reactor was a definite improvement and is still operating, the literature is filled with proponents of other reactor systems, each claiming its own advantages. This report provides a summary of the results of a study to compare the development potential of three of these reactor systems with the commercially operating Synthol-entrained bed reactor system. The commercial Synthol reactor is used as a benchmark against which the development potential of the other three reactors can be compared. Most of the information on which this study is based was supplied by the M.W. Kellogg Co. No information beyond that in the literature on the operation of the Synthol reactor system was available for consideration in preparing this study, nor were any details of the changes made to the original Synthol system to overcome the operating problems reported in the literature. Because of conflicting claims and results found in the literature, it was decided to concentrate a large part of this study on a kinetic analysis of the reactor systems, in order to provide a theoretical analysis of intrinsic strengths and weaknesses of the reactors unclouded by different catalysts, operating conditions and feed compositions. The remainder of the study considers the physical attributes of the four reactor systems and compares their respective investment costs, yields, catalyst requirements and thermal efficiencies from simplified conceptual designs.

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

    NASA Technical Reports Server (NTRS)

    Harty, Richard B.; Durand, Richard E.

    1993-01-01

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

  20. Lunar electric power systems utilizing the SP-100 reactor coupled to dynamic conversion systems. Final report

    SciTech Connect

    Harty, R.B.; Durand, R.E.

    1993-03-01

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

  1. Dual annular rotating [open quotes]windowed[close quotes] nuclear reflector reactor control system

    DOEpatents

    Jacox, M.G.; Drexler, R.L.; Hunt, R.N.M.; Lake, J.A.

    1994-03-29

    A nuclear reactor control system is provided in a nuclear reactor having a core operating in the fast neutron energy spectrum where criticality control is achieved by neutron leakage. The control system includes dual annular, rotatable reflector rings. There are two reflector rings: an inner reflector ring and an outer reflector ring. The reflectors are concentrically assembled, surround the reactor core, and each reflector ring includes a plurality of openings. The openings in each ring are capable of being aligned or non-aligned with each other. Independent driving means for each of the annular reflector rings is provided so that reactor criticality can be initiated and controlled by rotation of either reflector ring such that the extent of alignment of the openings in each ring controls the reflection of neutrons from the core. 4 figures.

  2. AUTOMATIC CONTROL OF T7 TANKER BOILING WATER REACTOR PROPULSION SYSTEM. PRELIMINARY DESIGN AND ECONOMIC EVALUATION

    Microsoft Academic Search

    R. B. Rice; J. I. Owens; W. M. Gaines; R. C. Larsen; R. J. Noorda

    1960-01-01

    The results of a technical and economie analy'sis of automatic ; propulsion system control as a possible design improvement in the direct cycle ; boiling water reactor propulsion system in a T7 tanker are presented. The ; technical feasibility of attaining a completely automated marine boiling water ; propulsion system was determined. Economic incentives for automation were ; evaluated. A

  3. Development of a measurement system of gap between CSB and RV to shorten a nuclear reactor installation period

    Microsoft Academic Search

    Do-Young Ko; Jae-Gon Lee; Yong-Chul Kang; Sung-Hwan Kim

    2009-01-01

    To decrease the timescale for the installation of a nuclear reactor, a reduced-scale model system is proposed for Korea's third generation nuclear reactor, the APR1400 (advanced power reactor 1400). The construction period of a nuclear power plant is one of the most important factors to make a company competitive in international nuclear energy markets. Our study is related to the

  4. Neutronics investigation of advanced self-cooled liquid blanket systems in the helical reactor

    Microsoft Academic Search

    T. Tanaka; A. Sagara; T. Muroga; M. Z. Youssef

    2008-01-01

    Neutronics investigations have been conducted in the design activity of the helical-type reactor Force Free Helical Reactor (FFHR2) adopting Flibe-cooled and Li-cooled advanced liquid blanket systems. In this study, comprehensive investigations and geometry modifications related to the tritium breeding ratios (TBRs), neutron shielding performance and neutron wall loading on the first walls in FFHR2 have been performed by improving the

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

    NASA Technical Reports Server (NTRS)

    George, Jeffrey A.

    1991-01-01

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

  6. Results of theoretical and experimental studies of hydrodynamics of circulation loops in circulating fluidized bed reactors and systems with interconnected reactors

    NASA Astrophysics Data System (ADS)

    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.

  7. Destruction of chlorobenzene and carbon tetrachloride in a two-stage molten salt oxidation reactor system.

    PubMed

    Yang, Hee-Chul; Cho, Yong-Jun; Eun, Hee-Chul; Kim, Eung-Ho

    2008-08-01

    Molten salt oxidation (MSO) is one of the promising alternative destruction technologies for chlorinated organics, because it is capable of trapping chlorine during organic destruction. This study investigated the characteristics of a two-stage MSO reactor system for the destruction of CCl(4) and C(6)H(5)Cl. Investigated parameters were the MSO reactor temperature (from 1023 K to 1223 K) and the excess oxidizing air feed rate (50% and 100%). The destruction of chlorinated solvents is substantial in the Li(2)CO(3)-Na(2)CO(3) eutectic molten salt, irrespective of the tested condition. However, further oxidation of CO, which is found to be the major destruction product, is not substantial due to the limited temperature and gas residence time in the MSO reactor. Increases in the reactor temperature as well as those in the oxidizing air feed rate consistently lead to decreased emissions of carbon monoxide. No significant influence of the MSO reactor operating condition on the chlorine capturing efficiency was found. Over 99.95% and 99.997% of the chlorine was captured in the hot MSO reactors during the C(6)H(5)Cl and CCl(4) destructions, respectively. This result suggests a relatively low potential of the MSO system in the recombination of chlorinated organics, when compared to a conventional incineration system. PMID:18501405

  8. 2014 APSECA Safe Programming Guidance of Function Block Diagram for Reactor Protection Systems Dong-Ah Lee*, Junbeom Yoo

    E-print Network

    in safety critical systems ­ Nuclear power plant · RPS (Reactor Protection System) · ESF-CCS (Engineering of Function Block Diagram for Reactor Protection Systems Software development in the nuclear power plant Nuclear Instrumentation & Control System R&D Center (KNICS) · FBD and Ladder Diagram (LD) to design

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

    SciTech Connect

    El-Genk, Mohamed; Tournier, Jean-Michel [Institute for Space and Nuclear Power Studies, University of New Mexico, Albuquerque, NM 87131 (United States); Chemical and Nuclear Engineering Department, University of New Mexico, Albuquerque, NM 87131 (United States)

    2005-02-06

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

  10. Air-lift reactor system for the treatment of waste-gas-containing monochlorobenzene.

    PubMed

    Joshi, Pradnya R; Deshmukh, Sharvari C; Morone, Amruta P; Kanade, Gajanan; Pandey, R A

    2013-01-01

    An air-lift bioreactor (ALR) system, applied for the treatment of waste-gas-containing monochlorobenzene (MCB) was seeded with pure culture of Acinetobacter calcoaceticus, isolated from soil as a starter seed. It was found that MCB was biologically converted to chloride as chloride was mineralized in the ALR. After the built up of the biomass in the ALR, the reactor parameters which have major influence on the removal efficiency and elimination capacity were studied using response surface methodology. The data generated by running the reactor for 150 days at varying conditions were fed to the model with a target to obtain the removal efficiency above 95% and the elimination capacity greater than 60%. The data analysis indicated that inlet loading was the major parameter affecting the elimination capacity and removal efficiency of >95%. The reactor when operated at optimized conditions resulted in enhanced performance of the reactor. PMID:24617061

  11. Enhancing VHTR passive safety and economy with thermal radiation based direct reactor auxiliary cooling system

    SciTech Connect

    Zhao, H.; Zhang, H.; Zou, L. [Idaho National Laboratory (United States); Sun, X. [Ohio State Univ. (United States)

    2012-07-01

    One of the most important requirements for Gen. IV Very High Temperature Reactor (VHTR) is passive safety. Currently all the gas cooled version of VHTR designs use Reactor Vessel Auxiliary Cooling System (RVACS) for passive decay heat removal. The RVACS can be characterized as a surface-based decay heat removal system. It is especially suitable for smaller power reactors since small systems have relatively larger surface area to volume ratio. However, RVACS limits the maximum achievable power level for modular VHTRs due to the mismatch between the reactor power (proportional to the core volume) and decay heat removal capability (proportional to the vessel surface area). Besides the safety considerations, VHTRs also need to be economical in order to compete with other reactor concepts and other types of energy sources. The limit of decay heat removal capability set by using RVACS has affected the economy of VHTRs. A potential alternative solution is to use a volume-based passive decay heat removal system, called Direct Reactor Auxiliary Cooling Systems (DRACS), to remove or mitigate the limitation on decay heat removal capability. DRACS composes of natural circulation loops with two sets of heat exchangers, one on the reactor side and another on the environmental side. For the reactor side, cooling pipes will be inserted into holes made in the outer or inner graphite reflector blocks. There will be gaps or annular regions formed between these cooling pipes and their corresponding surrounding graphite surfaces. Graphite has an excellent heat conduction property. By taking advantage of this feature, we can have a volume-based method to remove decay heat. The scalability can be achieved, if needed, by employing more rows of cooling pipes to accommodate higher decay heat rates. Since heat can easily conduct through the graphite regions among the holes made for the cooling pipes, those cooling pipes located further away from the active core region can still be very effective in removing decay heat. By removing the limit on the decay heat removal capability due to the limited available surface area as in a RVACS, the reactor power density and therefore the reactor power can be significantly increased, without losing the passive heat removal feature. This paper introduces the concept of using DRACS to enhance VHTR passive safety and economics. Three design options with different cooling pipe locations are discussed. Analysis results from a lumped volume based model and CFD simulations are presented. (authors)

  12. The muon system of the Daya Bay Reactor antineutrino experiment

    E-print Network

    Daya Bay Collaboration

    2014-11-28

    The Daya Bay experiment consists of functionally identical antineutrino detectors immersed in pools of ultrapure water in three well-separated underground experimental halls near two nuclear reactor complexes. These pools serve both as shields against natural, low-energy radiation, and as water Cherenkov detectors that efficiently detect cosmic muons using arrays of photomultiplier tubes. Each pool is covered by a plane of resistive plate chambers as an additional means of detecting muons. Design, construction, operation, and performance of these muon detectors are described.

  13. The Muon System of the Daya Bay Reactor Antineutrino Experiment

    DOE PAGESBeta

    An, F. P.; Hackenburg, R. W.; Brown, R. E.; Chasman, C.; Dale, E.; Diwan, M. V.; Gill, R.; Hans, S.; Isvan, Z.; Jaffe, D. E.; Kettell, S. H.; Littenberg, L.; Pearson, C. E.; Qian, X.; Theman, H.; Viren, B.; Worcester, E.; Yeh, M.; Zhang, C.

    2015-02-01

    The Daya Bay experiment consists of functionally identical antineutrino detectors immersed in pools of ultrapure water in three well-separated underground experimental halls near two nuclear reactor complexes. These pools serve both as shields against natural, low-energy radiation, and as water Cherenkov detectors that efficiently detect cosmic muons using arrays of photomultiplier tubes. Each pool is covered by a plane of resistive plate chambers as an additional means of detecting muons. Design, construction, operation, and performance of these muon detectors are described. (auth)

  14. The muon system of the Daya Bay Reactor antineutrino experiment

    NASA Astrophysics Data System (ADS)

    An, F. P.; Balantekin, A. B.; Band, H. R.; Beriguete, W.; Bishai, M.; Blyth, S.; Brown, R. E.; Butorov, I.; Cao, G. F.; Cao, J.; Carr, R.; Chan, Y. L.; Chang, J. F.; Chang, L.; Chang, Y.; Chasman, C.; Chen, H. S.; Chen, H. Y.; Chen, Q. Y.; Chen, S. J.; Chen, S. M.; Chen, X. C.; Chen, X. H.; Chen, Y.; Chen, Y. X.; Cheng, Y. P.; Cherwinka, J. J.; Chu, M. C.; Cummings, J. P.; Dale, E.; de Arcos, J.; Deng, Z. Y.; Ding, Y. Y.; Diwan, M. V.; Draeger, E.; Du, X. F.; Dwyer, D. A.; Edwards, W. R.; Ely, S. R.; Fu, J. Y.; Ge, L. Q.; Gill, R.; Goett, J.; Gonchar, M.; Gong, G. H.; Gong, H.; Gu, W. Q.; Guan, M. Y.; Guo, X. H.; Hackenburg, R. W.; Han, G. H.; Hans, S.; He, M.; He, Q.; Heeger, K. M.; Heng, Y. K.; Hinrichs, P.; Hor, Y. K.; Hsiung, Y. B.; Hu, B. Z.; Hu, L. J.; Hu, L. M.; Hu, T.; Hu, W.; Huang, E. C.; Huang, H. X.; Huang, H. Z.; Huang, X. T.; Huber, P.; Hussain, G.; Isvan, Z.; Jaffe, D. E.; Jaffke, P.; Jetter, S.; Ji, X. L.; Ji, X. P.; Jiang, H. J.; Jiao, J. B.; Johnson, R. A.; Kang, L.; Kebwaro, J. M.; Kettell, S. H.; Kramer, M.; Kwan, K. K.; Kwok, M. W.; Kwok, T.; Lai, W. C.; Lai, W. H.; Lau, K.; Lebanowski, L.; Lee, J.; Lei, R. T.; Leitner, R.; Leung, A.; Leung, J. K. C.; Lewis, C. A.; Li, D. J.; Li, F.; Li, G. S.; Li, Q. J.; Li, W. D.; Li, X. N.; Li, X. Q.; Li, Y. Z. B.; Liang, H.; Lin, C. J.; Lin, G. L.; Lin, P. Y.; Lin, S. K.; Link, J. M.; Littenberg, L.; Littlejohn, B. R.; Liu, D. W.; Liu, H.; Liu, J. C.; Liu, J. L.; Liu, S. S.; Liu, Y. B.; Lu, C.; Lu, H. Q.; Luk, K. B.; Ma, Q. M.; Ma, X. B.; Ma, X. Y.; Ma, Y. Q.; McDonald, K. T.; McFarlane, M. C.; McKeown, R. D.; Meng, Y.; Mitchell, I.; Mohapatra, D.; Morgan, J. E.; Nakajima, Y.; Napolitano, J.; Naumov, D.; Naumova, E.; Nemchenok, I.; Newsom, C.; Ngai, H. Y.; Ngai, W. K.; Ning, Z.; Ochoa-Ricoux, J. P.; Olshevski, A.; Patton, S.; Pec, V.; Pearson, C. E.; Peng, J. C.; Piilonen, L. E.; Pinsky, L.; Pun, C. S. J.; Qi, F. Z.; Qi, M.; Qian, X.; Raper, N.; Ren, B.; Ren, J.; Rosero, R.; Roskovec, B.; Ruan, X. C.; Shao, B. B.; Steiner, H.; Sun, G. X.; Sun, J. L.; Tam, Y. H.; Tang, X.; Themann, H.; Tsang, K. V.; Tsang, R. H. M.; Tull, C. E.; Tung, Y. C.; Viren, B.; Virostek, S.; Vorobel, V.; Wang, C. H.; Wang, L. S.; Wang, L. Y.; Wang, L. Z.; Wang, M.; Wang, N. Y.; Wang, R. G.; Wang, W.; Wang, W. W.; Wang, X.; Wang, Y. F.; Wang, Z.; Wang, Z.; Wang, Z. M.; Webber, D. M.; Wei, H. Y.; Wei, Y. D.; Wen, L. J.; Whisnant, K.; White, C. G.; Whitehead, L.; Wilhelmi, J.; Wise, T.; Wong, H. L. H.; Wong, S. C. F.; Worcester, E.; Wu, Q.; Xia, D. M.; Xia, J. K.; Xia, X.; Xing, Z. Z.; Xu, G. H.; Xu, J.; Xu, J. L.; Xu, J. Y.; Xu, Y.; Xue, T.; Yan, J.; Yang, C. G.; Yang, L.; Yang, M. S.; Yang, M. T.; Ye, M.; Yeh, M.; Yeh, Y. S.; Young, B. L.; Yu, G. Y.; Yu, J. Y.; Yu, Z. Y.; Zang, S. L.; Zhan, L.; Zhang, C.; Zhang, F. H.; Zhang, J. W.; Zhang, K.; Zhang, Q. M.; Zhang, S. H.; Zhang, Y. H.; Zhang, Y. M.; Zhang, Y. X.; Zhang, Z. J.; Zhang, Z. P.; Zhang, Z. Y.; Zhao, J.; Zhao, Q. W.; Zhao, Y.; Zhao, Y. B.; Zheng, L.; Zhong, W. L.; Zhou, L.; Zhou, Z. Y.; Zhuang, H. L.; Zou, J. H.

    2015-02-01

    The Daya Bay experiment consists of functionally identical antineutrino detectors immersed in pools of ultrapure water in three well-separated underground experimental halls near two nuclear reactor complexes. These pools serve both as shields against natural, low-energy radiation, and as water Cherenkov detectors that efficiently detect cosmic muons using arrays of photomultiplier tubes. Each pool is covered by a plane of resistive plate chambers as an additional means of detecting muons. Design, construction, operation, and performance of these muon detectors are described.

  15. Evaluation of a passive containment cooling system for a simplified BWR (boiling water reactor)

    Microsoft Academic Search

    J. Otonari; K. Arai; H. Oikawa; H. Nagasaka

    1989-01-01

    Simplified boiling water reactors (BWRs) are characterized for the adoption of a passive containment cooling system (PCCS) and a passive emergency core cooling system (ECCS). TOSPAC, which had been developed as the preliminary design code for several PCCS concepts, was compared with TRAC for verification. TOSPAC analyses were also performed to show the effectiveness of the isolation condenser (IC) as

  16. Boiling water reactor dynamics identification by the dynamic data system methodology

    Microsoft Academic Search

    Z. P. Luo; S. M. Wu

    1986-01-01

    A simplified model describing the complicated boiling water reactor (BWR) kinetics is developed by combining two different approaches: the analytical and the experimental, the latter based on the dynamic data system methodology. This model is capable of adequately representing the BWR main dynamics, which can be, in turn, conveniently used to predict its kinetics. Although the system exhibits some highly

  17. Wastewater Effluent Polishing Systems of Anaerobic Baffled Reactor Treating Black-water from Households

    E-print Network

    Richner, Heinz

    Wastewater Effluent Polishing Systems of Anaerobic Baffled Reactor Treating Black-water from of different integrated low-cost wastewater treatment systems, comprising one ABR as first treatment step filter and a vertical flow constructed wetland. A mixture of septage and domestic wastewater was used

  18. A microprocessor based monitoring system for a small nuclear reactor facility

    Microsoft Academic Search

    G. E. Miller; C. F. DeKeyser

    1980-01-01

    An inexpensive microprocessor based system has been designed and constructed for our 250 kilowatt TRIGA reactor facility. The system, which is beginning operational testing, can monitor on a continuous basis the status of up to 54 devices and maintain a record of events. These devices include fixed radiation monitors, pool water level trips, security alarms and an access control unit.

  19. Robotized system for retrieving fallen objects within the reactor vessel of a nuclear power plant (PWR)

    Microsoft Academic Search

    A. Iborra; B. Alvarez; P. J. Navarro; J. A. Pastor-Franco

    2000-01-01

    This paper presents an original teleoperated and robotized system (TRON) designed for retrieving foreign objects within lower internals of the reactor vessels at nuclear power plants (PWR). For performing these operations, the system does not require that the lower internals have to be retrieved or the fuel assemblies unloaded. The remote handler device is an articulated pole for accessing the

  20. Design and replacement of the Ohio State University research reactor's safety system

    Microsoft Academic Search

    Hatch

    1989-01-01

    The Ohio State University research reactor's (OSURR's) safety system was based on an electronic design of an electronic design of the 1950s. Due to difficulty in obtaining parts and excess downtime, it was decided to upgrade this system. Two avenues were pursued, each with a common goal of enhanced reliability and avoiding changes that would constitute unreviewed safety questions (pursuant

  1. Nuclear reactor system study for NASA/JPL

    NASA Technical Reports Server (NTRS)

    Palmer, R. G.; Lundberg, L. B.; Keddy, E. S.; Koenig, D. R.

    1982-01-01

    Reactor shielding, safety studies, and heat pipe development work are described. Monte Carlo calculations of gamma and neutron shield configurations show that substantial weight penalties are incurred if exposure at 25 m to neutrons and gammas must be limited to 10 to the 12th power nvt and 10 to the 6th power rad, instead of the 10 to the 13th power nvt and 10 to the 7th power rad values used earlier. For a 1.6 MW sub t reactor, the required shield weight increases from 400 to 815 kg. Water immersion critically calculations were extended to study the effect of water in fuel void spaces as well as in the core heat pipes. These show that the insertion into the core of eight blades of B4C with a mass totaling 2.5 kg will guarantee subcriticality. The design, fabrication procedure, and testing of a 4m long molybdenum/lithium heat pipe are described. It appears that an excess of oxygen in the wick prevented the attainment of expected performance capability.

  2. Laser anemometry measurements of natural circulation flow in a scale model PWR reactor system. [Pressurized Water Reactor

    NASA Technical Reports Server (NTRS)

    Kadambi, J. R.; Schneider, S. J.; Stewart, W. A.

    1986-01-01

    The natural circulation of a single phase fluid in a scale model of a pressurized water reactor system during a postulated grade core accident is analyzed. The fluids utilized were water and SF6. The design of the reactor model and the similitude requirements are described. Four LDA tests were conducted: water with 28 kW of heat in the simulated core, with and without the participation of simulated steam generators; water with 28 kW of heat in the simulated core, with the participation of simulated steam generators and with cold upflow of 12 lbm/min from the lower plenum; and SF6 with 0.9 kW of heat in the simulated core and without the participation of the simulated steam generators. For the water tests, the velocity of the water in the center of the core increases with vertical height and continues to increase in the upper plenum. For SF6, it is observed that the velocities are an order of magnitude higher than those of water; however, the velocity patterns are similar.

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

    NASA Technical Reports Server (NTRS)

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

    2009-01-01

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

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

    NASA Technical Reports Server (NTRS)

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

    2010-01-01

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

  5. Treatment of phthalic waste by anaerobic hybrid reactor

    SciTech Connect

    Tur, M.Y.; Huang, J.C. [Univ. of Missouri, Rolla, MO (United States)

    1997-11-01

    The anaerobic treatment performance of phthalic acid at 4,000 mg/L (dry weight) by a hybrid reactor consisting of an upflow anaerobic sludge blanket (UASB) and a biofilter was examined. Using anaerobic sewage sludge as the seed and glucose as a carbon supplement, it took 3 months to initiate phthalate degradation. After that, the glucose supplement could be discontinued. At 35 C and a phthalic loading of 20 g-COD/L-d, the chemical oxygen demand (COD) removal efficiency was nearly 95%. About 89.5% of the removed phthalic COD was converted to methane. When the phthalic loadings were increased to 26.7, 33.0, 39.7, and 46.3 g-COD/L-d, the COD removal efficiencies were progressively reduced to 78, 65, 58, and 47.7%, respectively. More than 95% of the residual effluent COD was composed of nondecomposed phthalic acid. In the hybrid reactor, 86% of the biomass was found in the UASB section while the remaining 14% was found in the biofilter section. The anaerobic sludge could lead to granulation. At 35 C and a phthalic loading of 26 g-COD/L-D, the overall specific removal rate was 0.81--0.85 g-COD/g VSS-d, and the corresponding methane production rate was 0.24--0.26 L CH{sub 4}/g VSS-d.

  6. The results of systems tests of the 500 kV busbar controllable shunting reactor in the Tavricheskaya substation

    SciTech Connect

    Gusev, S. I. [JSC 'FSK EES' (Russian Federation); Karpov, V. N.; Kiselev, A. N.; Kochkin, V. I. [Scientific-Research Institute of Electric Power Engineering (VNIIE) - Branch of the JSC 'NTTs Elektroenergetiki', (Russian Federation)

    2009-09-15

    The results of systems tests of the 500 kV busbar magnetization-controllable shunting reactor (CSR), set up in the Tavricheskaya substation, including measurements of the quality of the electric power, the harmonic composition of the network currents of the reactor for different values of the reactive power consumed, the determination of the regulating characteristics of the reactor, the speed of response of the shunting reactor in the current and voltage stabilization modes, and also the operation of the reactor under dynamic conditions for different perturbations, are presented. The results obtained are analyzed.

  7. Operation of Fusion Reactors in One Atmosphere of Air Instead of Vacuum Systems

    NASA Astrophysics Data System (ADS)

    Roth, J. Reece

    2009-07-01

    Engineering design studies of both magnetic and inertial fusion power plants have assumed that the plasma will undergo fusion reactions in a vacuum environment. Operation under vacuum requires an expensive additional major system for the reactor-a vacuum vessel with vacuum pumping, and raises the possibility of sudden unplanned outages if the vacuum containment is breached. It would be desirable in many respects if fusion reactors could be made to operate at one atmosphere with air surrounding the plasma, thus eliminating the requirement of a pressure vessel and vacuum pumping. This would have obvious economic, reliability, and engineering advantages for currently envisaged power plant reactors; it would make possible forms of reactor control not possible under vacuum conditions (i.e. adiabatic compression of the fusion plasma by increasing the pressure of surrounding gas); it would allow reactors used as aircraft engines to operate as turbojets or ramjets in the atmosphere, and it would allow reactors used as fusion rockets to take off from the surface of the earth instead of low earth orbit.

  8. Utilizing a Russian space nuclear reactor for a United States space mission: Systems integration issues

    SciTech Connect

    Reynolds, E.; Schaefer, E. [Johns Hopkins Univ., Laurel, MD (United States). Applied Physics Lab.; Polansky, G.; Lacy, J. [Phillips Lab., Albuquerque, NM (United States); Bocharov, A. [GDBMB, St. Petersburg (Russian Federation)

    1993-09-30

    The Nuclear Electric Propulsion Space Test Program (NEPSTP) has developed a cooperative relationship with several institutes of the former Soviet Union to evaluate Russian space hardware on a US spacecraft One component is the Topaz II Nuclear Power System; a built and flight qualified nuclear reactor that has yet to be tested in space. The access to the Topaz II reactor provides the NEPSTP with a rare opportunity; to conduct an early flight demonstration of nuclear electric propulsion at a relatively low cost. This opportunity, however, is not without challenges. Topaz II was designed to be compatible with Russian spacecraft and launch vehicles. It was manufactured and flight qualified by Russian techniques and standards and conforms to safety requirements of the former Soviet Union, not the United States. As it is desired to make minimal modifications to the Topaz II, integrating the reactor system with a United States spacecraft and launch vehicle presents an engineering challenge. This paper documents the lessons teamed regarding the integration of reactor based spacecraft and also some insight about integrating Russian hardware. It examines the planned integration flow along with specific reactor requirements that affect the spacecraft integration including American-Russian space system compatibility.

  9. System Analysis for Decay Heat Removal in Lead-Bismuth Cooled Natural Circulated Reactors

    SciTech Connect

    Takaaki Sakai; Yasuhiro Enuma [O-arai Engineering Center, Japan Nuclear Cycle Development Institute, Narita 4002, O-arai, Ibaraki, 311-1393 (Japan); Takashi Iwasaki [Nuclear Energy System Inc. Narita 4002, O-arai, Ibaraki, 311-1313 (Japan); Kazuhiro Ohyama [Advanced Reactor Technology Co. Ltd., 15-1 Tomihisa-cho, Shinjuku-ku, Tokyo, 162-0067 (Japan)

    2002-07-01

    Decay heat removal analyses for lead-bismuth cooled natural circulation reactors are described in this paper. A combined multi-dimensional plant dynamics code (MSG-COPD) has been developed to conduct the system analysis for the natural circulation reactors. For the preliminary study, transient analysis has been performed for a 100 MWe lead-bismuth-cooled reactor designed by Argonne National Laboratory (ANL). In addition, decay heat removal characteristics of a 400 MWe lead-bismuth-cooled natural circulation reactor designed by Japan Nuclear Cycle Development Institute (JNC) has been evaluated by using MSG-COPD. PRACS (Primary Reactor Auxiliary Cooling System) is prepared for the JNC's concept to get sufficient heat removal capacity. During 2000 sec after the transient, the outlet temperature shows increasing tendency up to the maximum temperature of 430 Centigrade, because the buoyancy force in a primary circulation path is temporary reduced. However, the natural circulation is recovered by the PRACS system and the out let temperature decreases successfully. (authors)

  10. System Analysis for Decay Heat Removal in Lead-Bismuth-Cooled Natural-Circulation Reactors

    SciTech Connect

    Sakai, Takaaki [Japan Nuclear Cycle Development Institute (Japan); Enuma, Yasuhiro [Japan Nuclear Cycle Development Institute (Japan); Iwasaki, Takashi [Nuclear Energy System, Inc. (Japan)

    2004-03-15

    Decay heat removal analyses for lead-bismuth-cooled natural-circulation reactors are described in this paper. A combined multidimensional plant dynamics code (MSG-COPD) has been developed to conduct the system analysis for the natural-circulation reactors. For the preliminary study, transient analysis has been performed for a 300-MW(thermal) lead-bismuth-cooled reactor designed by Argonne National Laboratory. In addition, decay heat removal characteristics of a 400-MW(electric) lead-bismuth-cooled natural-circulation reactor designed by the Japan Nuclear Cycle Development Institute (JNC) has been evaluated by using MSG-COPD. The primary reactor auxiliary cooling system (PRACS) is prepared for the JNC concept to get sufficient heat removal capacity. During 2000 s after the transient, the outlet temperature shows increasing tendency up to the maximum temperature of 430 deg. C because the buoyancy force in a primary circulation path is temporarily reduced. However, the natural circulation is recovered by the PRACS system, and the outlet temperature decreases successfully.

  11. The effects of aging on Boiling Water Reactor core isolation cooling system

    SciTech Connect

    Lee, Bom Soon

    1994-06-01

    A study was performed to assess the effects of aging on the Reactor Core Isolation Cooling system in commercial Boiling Water Reactors. This study is part of the Nuclear Plant Aging Research program sponsored by the US Nuclear Regulatory Commission. The failure data, from national databases, as well as plant specific data were reviewed and analyzed to understand the effects of aging on the RCIC system. This analysis identified important components that should receive the highest priority in terms of aging management. The aging characterization provided information on the effects of aging on component failure frequency, failure modes, and failure causes.

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

    NASA Technical Reports Server (NTRS)

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

    1991-01-01

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

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

    SciTech Connect

    Pablo Rubiolo, Principal Investigator

    2003-03-21

    The main features of this project were the development of a long life (up to 10 years) Liquid Metal Reactor (LMR) and a static conversion subsystem comprising an Alkali Metal Thermal-to-Electric (AMTEC) topping cycle and a ThermoElectric (TE) Bottom cycle. Various coupling options of the LMR with the energy conversion subsystem were explored and, base in the performances found in this analysis, an Indirect Coupling (IC) between the LMR and the AMTEC/TE converters with Alkali Metal Boilers (AMB) was chosen as the reference design. The performance model of the fully integrated sodium-and potassium-AMTEC/TE converters shows that a combined conversion efficiency in excess of 30% could be achieved by the plant. (B204)

  14. Novel duplex vapor: Electrochemical method for silicon solar cells. [chemical reactor for a silicon sodium reaction system

    NASA Technical Reports Server (NTRS)

    Nanis, L.; Sanjurjo, A.; Sancier, K.

    1979-01-01

    The scaled up chemical reactor for a SiF4-Na reaction system is examined for increased reaction rate and production rate. The reaction system which now produces 5 kg batches of mixed Si and NaF is evaluated. The reactor design is described along with an analysis of the increased capacity of the Na chip feeder. The reactor procedure is discussed and Si coalescence in the reaction products is diagnosed.

  15. Reactor control system upgrade for the McClellan Nuclear Radiation Center Sacramento, CA.

    SciTech Connect

    Power, M. A.

    1999-03-10

    Argonne National Laboratory is currently developing a new reactor control system for the McClellan Nuclear Radiation Facility. This new control system not only provides the same functionality as the existing control system in terms of graphic displays of reactor process variables, data archival capability, and manual, automatic, pulse and square-wave modes of operation, but adds to the functionality of the previous control system by incorporating signal processing algorithms for the validation of sensors and automatic calibration and verification of control rod worth curves. With the inclusion of these automated features, the intent of this control system is not to replace the operator but to make the process of controlling the reactor easier and safer for the operator. For instance, an automatic control rod calibration method reduces the amount of time to calibrate control rods from days to minutes, increasing overall reactor utilization. The control rod calibration curve, determined using the automatic calibration system, can be validated anytime after the calibration, as long as the reactor power is between 50W and 500W. This is done by banking all of the rods simultaneously and comparing the tabulated rod worth curves with a reactivity computer estimate. As long as the deviation between the tabulated values and the reactivity estimate is within a prescribed error band, then the system is in calibration. In order to minimize the amount of information displayed, only the essential flux-related data are displayed in graphical format on the control screen. Information from the sensor validation methods is communicated to the operators via messages, which appear in a message window. The messages inform the operators that the actual process variables do not correlate within the allowed uncertainty in the reactor system. These warnings, however, cannot cause the reactor to shutdown automatically. The reactor operator has the ultimate responsibility of using this information to either keep the reactor operating or to shut the reactor down. In addition to new developments in the signal processing realm, the new control system will be migrating from a PC-based computer platform to a Sun Solaris-based computer platform. The proven history of stability and performance of the Sun Sohuis operating system are the main advantages to this change. The I/O system will also be migrating from a PC-based data collection system, which communicates plant data to the control computer using RS-232 connections, to an Ethernet-based I/O system. The Ethernet Data Acquisition System (EDAS) modules from Intelligent Instrumentation, Inc. provide an excellent solution for embedded control of a system using the more universally-accepted data transmission standard of TCP/IP. The modules contain a PROM, which operates all of the functionality of the I/O module, including the TCP/IP network access. Thus the module does not have an internal, sophisticated operating system to provide functionality but rather a small set hard-coded of instructions, which almost eliminates the possibility of the module failing due to software problems. An internal EEPROM can be modified over the Internet to change module configurations. Once configured, the module is contacted just like any other Internet host using TCP/IP socket calls. The main advantage to this architecture is its flexibility, expandability, and high throughput.

  16. Enhancing VHTR Passive Safety and Economy with Thermal Radiation Based Direct Reactor Auxiliary Cooling System

    SciTech Connect

    Haihua Zhao; Hongbin Zhang; Ling Zou; Xiaodong Sun

    2012-06-01

    One of the most important requirements for Gen. IV Very High Temperature Reactor (VHTR) is passive safety. Currently all the gas cooled version of VHTR designs use Reactor Vessel Auxiliary Cooling System (RVACS) for passive decay heat removal. The decay heat first is transferred to the core barrel by conduction and radiation, and then to the reactor vessel by thermal radiation and convection; finally the decay heat is transferred to natural circulated air or water systems. RVACS can be characterized as a surface based decay heat removal system. The RVACS is especially suitable for smaller power reactors since small systems have relatively larger surface area to volume ratio. However, RVACS limits the maximum achievable power level for modular VHTRs due to the mismatch between the reactor power (proportional to volume) and decay heat removal capability (proportional to surface area). When the relative decay heat removal capability decreases, the peak fuel temperature increases, even close to the design limit. Annular core designs with inner graphite reflector can mitigate this effect; therefore can further increase the reactor power. Another way to increase the reactor power is to increase power density. However, the reactor power is also limited by the decay heat removal capability. Besides the safety considerations, VHTRs also need to be economical in order to compete with other reactor concepts and other types of energy sources. The limit of decay heat removal capability set by using RVACS has affected the economy of VHTRs. A potential alternative solution is to use a volume-based passive decay heat removal system, called Direct Reactor Auxiliary Cooling Systems (DRACS), to remove or mitigate the limitation on decay heat removal capability. DRACS composes of natural circulation loops with two sets of heat exchangers, one on the reactor side and another on the environment side. For the reactor side, cooling pipes will be inserted into holes made in the outer or inner graphite reflector blocks. There will be gaps between these cooling pipes and their corresponding surrounding graphite surfaces. Graphite has an excellent heat conduction property. By taking advantage of this feature, we can have a volume-based method to remove decay heat. The scalability can be achieved, if needed, by employing more rows of cooling pipes to accommodate higher decay heat rates. Since heat can easily conduct through the graphite regions between the holes made for the cooling pipes, those cooling pipes located further away from the active core region can still be very effective in removing decay heat. By removing the limit on the decay heat removal capability due to the limited available surface area as in a RVACS, the reactor power and power density can be significantly increased, without losing the passive heat removal feature. This paper will introduce the concept of using DRACS to enhance VHTR passive safety and economics. Three design options will be discussed, depending on the cooling pipe locations. Analysis results from a lumped volume based model and CFD simulations will be presented.

  17. Combustion flame-plasma hybrid reactor systems, and chemical reactant sources

    DOEpatents

    Kong, Peter C

    2013-11-26

    Combustion flame-plasma hybrid reactor systems, chemical reactant sources, and related methods are disclosed. In one embodiment, a combustion flame-plasma hybrid reactor system comprising a reaction chamber, a combustion torch positioned to direct a flame into the reaction chamber, and one or more reactant feed assemblies configured to electrically energize at least one electrically conductive solid reactant structure to form a plasma and feed each electrically conductive solid reactant structure into the plasma to form at least one product is disclosed. In an additional embodiment, a chemical reactant source for a combustion flame-plasma hybrid reactor comprising an elongated electrically conductive reactant structure consisting essentially of at least one chemical reactant is disclosed. In further embodiments, methods of forming a chemical reactant source and methods of chemically converting at least one reactant into at least one product are disclosed.

  18. Expert system connected to the University of Texas Research Reactor

    SciTech Connect

    Bauer, T.L.; Wehring, B.W. [Univ. of Texas, Austin (United States)

    1994-12-31

    A PC-based system for implementing and evaluating expert system strategies has been developed to operate with the UT-TRIGA. A primary purpose for the system development is to research methods of improving digital control system reliability. One proposal currently being investigated is the use of program functional diversity by applying functionally different programming techniques.

  19. Nuclear reactor with makeup water assist from residual heat removal system

    DOEpatents

    Corletti, M.M.; Schulz, T.L.

    1993-12-07

    A pressurized water nuclear reactor uses its residual heat removal system to make up water in the reactor coolant circuit from an in-containment refueling water supply during staged depressurization leading up to passive emergency cooling by gravity feed from the refueling water storage tank, and flooding of the containment building. When depressurization commences due to inadvertence or a manageable leak, the residual heat removal system is activated manually and prevents flooding of the containment when such action is not necessary. Operation of the passive cooling system is not impaired. A high pressure makeup water storage tank is coupled to the reactor coolant circuit, holding makeup coolant at the operational pressure of the reactor. The staged depressurization system vents the coolant circuit to the containment, thus reducing the supply of makeup coolant. The level of makeup coolant can be sensed to trigger opening of successive depressurization conduits. The residual heat removal pumps move water from the refueling water storage tank into the coolant circuit as the coolant circuit is depressurized, preventing reaching the final depressurization stage unless the makeup coolant level continues to drop. The residual heat removal system can also be coupled in a loop with the refueling water supply tank, for an auxiliary heat removal path. 2 figures.

  20. A Conceptual Multi-Megawatt System Based on a Tungsten CERMET Reactor

    SciTech Connect

    Jonathan A. Webb; Brian Gross

    2011-02-01

    Abstract. A conceptual reactor system to support Multi-Megawatt Nuclear Electric Propulsion is investigated within this paper. The reactor system consists of a helium cooled Tungsten-UN fission core, surrounded by a beryllium neutron reflector and 13 B4C control drums coupled to a high temperature Brayton power conversion system. Excess heat is rejected via carbon reinforced heat pipe radiators and the gamma and neutron flux is attenuated via segmented shielding consisting of lithium hydride and tungsten layers. Turbine inlet temperatures ranging from 1300 K to 1500 K are investigated for their effects on specific powers and net electrical outputs ranging from 1 MW to 100 MW. The reactor system is estimated to have a mass, which ranges from 15 Mt at 1 MWe and a turbine inlet temperature of 1500 K to 1200 Mt at 100 MWe and a turbine temperature of 1300 K. The reactor systems specific mass ranges from 32 kg/kWe at a turbine inlet temperature of 1300 K and a power of 1 MWe to 9.5 kg/kW at a turbine temperature of 1500 K and a power of 100 MWe.

  1. Nuclear reactor with makeup water assist from residual heat removal system

    DOEpatents

    Corletti, Michael M. (New Kensington, PA); Schulz, Terry L. (Murrysville, PA)

    1993-01-01

    A pressurized water nuclear reactor uses its residual heat removal system to make up water in the reactor coolant circuit from an in-containment refueling water supply during staged depressurization leading up to passive emergency cooling by gravity feed from the refueling water storage tank, and flooding of the containment building. When depressurization commences due to inadvertence or a manageable leak, the residual heat removal system is activated manually and prevents flooding of the containment when such action is not necessary. Operation of the passive cooling system is not impaired. A high pressure makeup water storage tank is coupled to the reactor coolant circuit, holding makeup coolant at the operational pressure of the reactor. The staged depressurization system vents the coolant circuit to the containment, thus reducing the supply of makeup coolant. The level of makeup coolant can be sensed to trigger opening of successive depressurization conduits. The residual heat removal pumps move water from the refueling water storage tank into the coolant circuit as the coolant circuit is depressurized, preventing reaching the final depressurization stage unless the makeup coolant level continues to drop. The residual heat removal system can also be coupled in a loop with the refueling water supply tank, for an auxiliary heat removal path.

  2. Assessing the influence of reactor system design criteria on the performance of model colon fermentation units.

    PubMed

    Moorthy, Arun S; Eberl, Hermann J

    2014-04-01

    Fermentation reactor systems are a key platform in studying intestinal microflora, specifically with respect to questions surrounding the effects of diet. In this study, we develop computational representations of colon fermentation reactor systems as a way to assess the influence of three design elements (number of reactors, emptying mechanism, and inclusion of microbial immobilization) on three performance measures (total biomass density, biomass composition, and fibre digestion efficiency) using a fractional-factorial experimental design. It was determined that the choice of emptying mechanism showed no effect on any of the performance measures. Additionally, it was determined that none of the design criteria had any measurable effect on reactor performance with respect to biomass composition. It is recommended that model fermentation systems used in the experimenting of dietary effects on intestinal biomass composition be streamlined to only include necessary system design complexities, as the measured performance is not benefited by the addition of microbial immobilization mechanisms or semi-continuous emptying scheme. Additionally, the added complexities significantly increase computational time during simulation experiments. It was also noted that the same factorial experiment could be directly adapted using in vitro colon fermentation systems. PMID:24216456

  3. SUSEE: A Compact, Lightweight Space Nuclear Power System Using Present Water Reactor Technology

    SciTech Connect

    Maise, George; Powell, James; Paniagua, John [Plus Ultra Technologies, Incorporated, Shoreham, NY 11786 (United States)

    2006-01-20

    The SUSEE space reactor system uses existing nuclear fuels and the standard steam cycle to generate electrical and thermal power for a wide range of in-space and surface applications, including manned bases, sub-surface mobile probes to explore thick ice deposits on Mars and the Jovian moons, and mobile rovers. SUSEE cycle efficiency, thermal to electric, ranges from {approx}20 to 24%, depending on operating parameters. Rejection of waste heat is by a lightweight condensing radiator that can be launched as a compact rolled-up package and deployed into flat panels when appropriate. The 50 centimeter diameter SUSEE reactor can provide power over the range of 10 kW(e) to 1 MW(e) for a period of 10 years. Higher power outputs are possible using slightly larger reactors. System specific weight (reactor, turbine, generator, piping, and radiator) is {approx}3 kg/kW(e). Two SUSEE reactor options are described, based on the existing Zr/O2 cermet and the UH3/ZrH2 TRIGA nuclear fuels.

  4. Study of reactor Brayton power systems for nuclear electric spacecraft

    NASA Technical Reports Server (NTRS)

    1979-01-01

    The feasibility of using Brayton power systems for nuclear electric spacecraft was investigated. The primary performance parameters of systems mass and radiator area were determined for systems from 100 to 1000 kW sub e. Mathematical models of all system components were used to determine masses and volumes. Two completely independent systems provide propulsion power so that no single-point failure can jeopardize a mission. The waste heat radiators utilize armored heat pipes to limit meteorite puncture. The armor thickness was statistically determined to achieve the required probability of survival. A 400 kW sub e reference system received primary attention as required by the contract. The components of this system were defined and a conceptual layout was developed with encouraging results. An arrangement with redundant Brayton power systems having a 1500 K (2240 F) turbine inlet temperature was shown to be compatible with the dimensions of the space shuttle orbiter payload bay.

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

    SciTech Connect

    Memmott, M. J.; Stansbury, C.; Taylor, C. [Westinghouse Electric Company LLC, 600 Cranberry Woods Drive, Cranberry Twp. PA 16066 (United States)

    2012-07-01

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

  6. A knowledge-based system for power maneuver planning for boiling water reactors

    Microsoft Academic Search

    M. Kinoshita; T. Fukuzaki; A. Nishimura; Y. Fukasawa; T. Matsuki

    1991-01-01

    This paper reports on a prototype knowledge-based system that assists plant site engineers in power maneuver planning for boiling water reactor (BWR) startup and load-following that has been developed. In the conventional method, engineers formulate these plans based on their own expertise and on core simulation programs that require long running times. To more quickly provide a suitable plan, the

  7. An autonomous long-term fast reactor system and the principal design limitations of the concept

    E-print Network

    Tsvetkova, Galina Valeryevna

    2004-09-30

    (STAR-H2) and represents further refinement of the STAR-H2 concept towards an economical, proliferation-resistant, sustainable, multi-purpose nuclear energy system. The development of the ALM-FR design has been performed considering this reactor within...

  8. Mechanistic modeling of heavy metal biosorption in batch and membrane reactor systems

    Microsoft Academic Search

    F. Pagnanelli; F. Beolchini; A. Esposito; L. Toro; F. Vegliò

    2003-01-01

    In this study, a microbial culture of Arthrobacter sp. was characterized and tested as heavy metal biosorbent in different experimental conditions (pH and biomass concentrations) and operative configurations (free cell in batch system and membrane reactor).Biosorption batch trials with free cells were carried out using an original procedure defined as “subsequent additions method” (SAM), consisting of successive additions of heavy

  9. Stochastic Analysis of Nonlinear Point Reactor Systems with Colored Multiplicative Noise

    Microsoft Academic Search

    Okitsugu SAKO

    1980-01-01

    This paper presents a master equation formulation of the fluctuations of neutron density in a nonlinear power reactor model system perturbed by colored multiplicative noise using the Van Kampen's master equation. The reactivity and power reactivity feedback coefficient are assumed to be stationary stochastic processes with short but finite correlation times.Some statistical characteristics of the neutron density fluctuations, such as

  10. Package Flow Model and Its Fuzzy Implementation for Simulating Nuclear Reactor System Dynamics

    Microsoft Academic Search

    Hiroshi MATSUOKA; Misako ISHIGURO

    1996-01-01

    A simple intuitive simulation model, which we call “Package Flow Model”, has been developed to evaluate physical processes in nuclear reactor system from a macroscopic point of view. In the previous paper, we showed the physical process of each energy generation and transfer stage in a PWR could be modeled by PFM, and its dynamics could be approximately simulated by

  11. FUNCTION POINT ANALYSIS: AN APPLICATION TO A NUCLEAR REACTOR PROTECTION SYSTEM

    Microsoft Academic Search

    Nihal Kececi; Ming Li; Carol Smidts

    1999-01-01

    This paper presents an application of full function point analysis to the estimation of the size of real -time control software. The full function point counting technique is briefly described. Its usage is illustrated on a part of the Westinghouse Reactor Protection Control System and the results analyzed. We further describe a technique for the graphical representation of requirements that

  12. Recurrent neuro-fuzzy system for fault detection and isolation in nuclear reactors

    Microsoft Academic Search

    Alexandre Evsukoff; Sylviane Gentil

    2005-01-01

    This paper presents an application of recurrent neuro-fuzzy systems to fault detection and isolation in nuclear reactors. A general framework is adopted, in which a fuzzification module is linked to an inference module that is actually a neural network adapted to the recognition of the dynamic evolution of process variables and related faults. Process data is fuzzified in order to

  13. Research article A modified batch reactor system to study equilibrium-reactive

    E-print Network

    Clement, Prabhakar

    Research article A modified batch reactor system to study equilibrium-reactive transport problems Engineering, Auburn University, Auburn, AL 36849, USA b School of Earth and Environmental Sciences, Seoul National University, 151-747, Republic of Korea a r t i c l e i n f o a b s t r a c t Article history

  14. Revealing systems of flaws in high-pressure reactors operating under dynamic pulse loading

    Microsoft Academic Search

    Gregory Muravin; Curd W. Adams; Boris Muravin; Luidmila Lezvinsky

    2004-01-01

    Quantitative Acoustic Emission (QAE) technology, physical and mathematical models were created for the reliable and precise identification and evaluation of the danger level (the J-integral value) of a developing main crack in a system of interacting micro- cracks, and the reliable assessment of the remaining lifetime of low density polyethylene (LDPE) reactor tubes that contain cracks. These innovations made it

  15. Development and Tests of the LED Calibration System for the Daya Bay Reactor Neutrino Experiment

    Microsoft Academic Search

    Melinda Morang

    2007-01-01

    The Daya Bay reactor neutrino experiment must measure the neutrino rate and spectrum with very precision. Thus, the detector modules must be carefully calibrated in order to produce reliable data. This study consists of hardware research and development for the LED portion of the detector calibration system, for which a fast timing resolution is key. We used a photomultiplier tube

  16. Development of optical components for in-vessel viewing systems used for fusion experimental reactor

    Microsoft Academic Search

    Kenjiro Obara; Satoshi Kakudate; Kiyoshi Oka; Eisuke Tada; Yosuke Morita; Masahiro Seki

    1994-01-01

    Optical components including imagefiber, periscope, glass, reflecting mirror and adhesive for lens are essential elements of in-vessel viewing system use for fusion experimental reactor and extensive of gamma irradiation tests have been conducted. These components were irradiated in the range of 1 MGy - 100 MGy under the average exposure dose rate of 1 X 106 R\\/h. As a result,

  17. Space reactor/Stirling cycle systems for high power Lunar applications

    SciTech Connect

    Schmitz, P.D. [Sverdrup Technology, Inc., Brook Park, OH (United States). Lewis Research Center Group; Mason, L.S. [National Aeronautics and Space Administration, Cleveland, OH (United States). Lewis Research Center

    1994-09-01

    NASA`s Space Exploration Initiative (SEI) has proposed the use of high power nuclear power systems on the lunar surface as a necessary alternative to solar power. Because of the long lunar night ({approximately} 14 earth days) solar powered systems with the requisite energy storage in the form of regenerative fuel cells or batteries becomes prohibitively heavy at high power levels ({approximately} 100 kWe). At these high power levels nuclear power systems become an enabling technology for variety of missions. One way of producing power on the lunar surface is with an SP-100 class reactor coupled with Stirling power converters. In this study, analysis and characterization of the SP-100 class reactor coupled with Free Piston Stirling Power Conversion (FPSPC) system will be performed. Comparison of results with previous studies of other systems, particularly Brayton and Thermionic, are made.

  18. Supervisory control design based on hybrid systems and fuzzy events detection. Application to an oxichlorination reactor.

    PubMed

    Altamiranda, Edmary; Torres, Horacio; Colina, Eliezer; Chacón, Edgar

    2002-10-01

    This paper presents a supervisory control scheme based on hybrid systems theory and fuzzy events detection. The fuzzy event detector is a linguistic model, which synthesizes complex relations between process variables and process events incorporating experts' knowledge about the process operation. This kind of detection allows the anticipation of appropriate control actions, which depend upon the selected membership functions used to characterize the process under scrutiny. The proposed supervisory control scheme was successfully implemented for an oxichlorination reactor in a vinyl monomer plant. This implementation has allowed improvement of reactor stability and reduction of raw material consumption. PMID:12398279

  19. SAFIRE: A systems analysis code for ICF (inertial confinement fusion) reactor economics

    SciTech Connect

    McCarville, T.J.; Meier, W.R.; Carson, C.F.; Glasgow, B.B.

    1987-01-12

    The SAFIRE (Systems Analysis for ICF Reactor Economics) code incorporates analytical models for scaling the cost and performance of several inertial confinement fusion reactor concepts for electric power. The code allows us to vary design parameters (e.g., driver energy, chamber pulse rate, net electric power) and evaluate the resulting change in capital cost of power plant and the busbar cost of electricity. The SAFIRE code can be used to identify the most attractive operating space and to identify those design parameters with the greatest leverage for improving the economics of inertial confinement fusion electric power plants.

  20. Flow-induced vibration and instability of some nuclear-reactor-system components. [PWR

    SciTech Connect

    Chen, S.S.

    1983-01-01

    The high-velocity coolant flowing through a reactor system component is a source of energy that can induce component vibration and instability. In fact, many reactor components have suffered from excessive vibration and/or dynamic instability. The potential for detrimental flow-induced vibration makes it necessary that design engineers give detailed considerations to the flow-induced vibration problems. Flow-induced-vibration studies have been performed in many countries. Significant progress has been made in understanding the different phenomena and development of design guidelines to avoid damaging vibration. The purpose of this paper is to present an overview of the recent progress in several selected areas, to discuss some new results and to indentify future research needs. Specifically, the following areas will be presented: examples of flow-induced-vibration problems in reactor components; excitation mechanisms and component response characteristics; instability mechanisms and stability criteria; design considerations; and future research needs.

  1. Development of a novel integrated continuous reactor system for biocatalytic production of biodiesel.

    PubMed

    Chattopadhyay, Soham; Sen, Ramkrishna

    2013-11-01

    A novel integrated immobilized enzyme-reactor system involving a continuous stirred tank reactor with two packed bed reactors in series was developed for the continuous production of biodiesel. The problem of methanol solubility into oil was solved by introducing a stirred tank reactor to dissolve methanol into partially converted oil. This step made the process perfectly continuous without requiring any organic solvent and intermittent methanol addition in the process. The substrate feeding rate of 0.74 mL/min and enzyme loading of 0.75 g per reactor were determined to be optimum for maximum biodiesel yield. The integrated continuous process was stable up to 45 cycles with biodiesel productivity of 137.2 g/L/h, which was approximately 5 times higher than solvent free batch process. In comparison with the processes reported in literature using expensive Novozyme 435 and hazardous organic solvent, the present process is completely green and perfectly continuous with economic and environmental advantages. PMID:24001564

  2. System simulation of a multicell thermionic space power reactor

    Microsoft Academic Search

    Alan Vincent von Arx

    1999-01-01

    For many years, thermionic power has been considered for space application. The prominent feature of the power conversion system is that there are no moving parts. Although designs have been developed by various organizations, no comprehensive system models are known to exist which can simulate transient behavior of a multicell design nor is there a method to directly couple these

  3. An autonomous long-term fast reactor system and the principal design limitations of the concept

    NASA Astrophysics Data System (ADS)

    Tsvetkova, Galina Valeryevna

    The objectives of this dissertation were to find a principal domain of promising and technologically feasible reactor physics characteristics for a multi-purpose, modular-sized, lead-cooled, fast neutron spectrum reactor fueled with an advanced uranium-transuranic-nitride fuel and to determine the principal limitations for the design of an autonomous long-term multi-purpose fast reactor (ALM-FR) within the principal reactor physics characteristic domain. The objectives were accomplished by producing a conceptual design for an ALM-FR and by analysis of the potential ALM-FR performance characteristics. The ALM-FR design developed in this dissertation is based on the concept of a secure transportable autonomous reactor for hydrogen production (STAR-H2) and represents further refinement of the STAR-H2 concept towards an economical, proliferation-resistant, sustainable, multi-purpose nuclear energy system. The development of the ALM-FR design has been performed considering this reactor within the frame of the concept of a self-consistent nuclear energy system (SCNES) that satisfies virtually all of the requirements for future nuclear energy systems: efficient energy production, safety, self-feeding, non-proliferation, and radionuclide burning. The analysis takes into consideration a wide range of reactor design aspects including selection of technologically feasible fuels and structural materials, core configuration optimization, dynamics and safety of long-term operation on one fuel loading, and nuclear material non-proliferation. Plutonium and higher actinides are considered as essential components of an advanced fuel that maintains long-term operation. Flexibility of the ALM-FR with respect to fuel compositions is demonstrated acknowledging the principal limitations of the long-term burning of plutonium and higher actinides. To ensure consistency and accuracy, the modeling has been performed using state-of-the-art computer codes developed at Argonne National Laboratory. As a result of the computational analysis performed in this work, the ALM-FR design provides for the possibility of continuous operation during about 40 years on one fuel loading containing mixture of depleted uranium with plutonium and higher actinides. All reactor physics characteristics of the ALM-FR are kept within technological limits ensuring safety of ultra-long autonomous operation. The results obtained provide for identification of physical features of the ALM-FR that significantly influence flexibility of the design and its applications. The special emphasis is given to existing limitations on the utilization of higher actinides as a fuel component.

  4. Digital control of research reactors

    Microsoft Academic Search

    J. C.. Crump; W. J. Richards; C. C. Heidel

    1991-01-01

    Research reactors provide an important service for the nuclear industry. Developments and innovations used for research reactors can be later applied to larger power reactors. Their relatively inexpensive cost allows research reactors to be an excellent testing ground for the reactors of tomorrow. One area of current interest is digital control of research reactor systems. Digital control systems offer the

  5. Startup thaw concept for the SP-100 space reactor power system

    NASA Technical Reports Server (NTRS)

    Kirpich, A.; Das, A.; Choe, H.; Mcnamara, E.; Switick, D.; Bhandari, P.

    1990-01-01

    A thaw concept for a space reactor power system which employs lithium as a circulant for both the heat-transport and the heat-rejection fluid loops is presented. An exemplary thermal analysis for a 100-kWe (i.e., SP-100) system is performed. It is shown that the design of the thaw system requires a thorough knowledge of the various physical states of the circulant throughout the system, both spatially and temporally, and that the design has to provide adequate margins for the system to avoid a structural or thermally induced damage.

  6. Hydraulic Shuttle Irradiation System (HSIS) Recently Installed in the Advanced Test Reactor (ATR)

    SciTech Connect

    A. Joseph Palmer; Gerry L. McCormick; Shannon J. Corrigan

    2010-06-01

    2010 International Congress on Advances in Nuclear Power Plants (ICAPP’10) ANS Annual Meeting Imbedded Topical San Diego, CA June 13 – 17, 2010 Hydraulic Shuttle Irradiation System (HSIS) Recently Installed in the Advanced Test Reactor (ATR) Author: A. Joseph Palmer, Mechanical Engineer, Irradiation Test Programs, 208-526-8700, Joe.Palmer@INL.gov Affiliation: Idaho National Laboratory P.O. Box 1625, MS-3840 Idaho Falls, ID 83415 INL/CON-10-17680 ABSTRACT Most test reactors are equipped with shuttle facilities (sometimes called rabbit tubes) whereby small capsules can be inserted into the reactor and retrieved during power operations. With the installation of Hydraulic Shuttle Irradiation System (HSIS) this capability has been restored to the Advanced Test Reactor (ATR) at Idaho National Laboratory (INL). The general design and operating principles of this system were patterned after the hydraulic rabbit at Oak Ridge National Laboratory’s (ORNL) High Flux Isotope Reactor (HFIR), which has operated successfully for many years. Using primary coolant as the motive medium the HSIS system is designed to simultaneously transport fourteen shuttle capsules, each 16 mm OD x 57 mm long, to and from the B-7 position of the reactor. The B-7 position is one of the higher flux positions in the reactor with typical thermal and fast (>1 Mev) fluxes of 2.8E+14 n/cm2/sec and 1.9E+14 n/cm2/sec respectively. The available space inside each shuttle is approximately 14 mm diameter x 50 mm long. The shuttle containers are made from titanium which was selected for its low neutron activation properties and durability. Shuttles can be irradiated for time periods ranging from a few minutes to several months. The Send and Receive Station (SRS) for the HSIS is located 2.5 m deep in the ATR canal which allows irradiated shuttles to be easily moved from the SRS to a wet loaded cask, or transport pig. The HSIS system first irradiated (empty) shuttles in September 2009 and has since completed a Readiness Assessment in November 2009. The HSIS is a key component of the ATR National Scientific User Facility (NSUF) operated by Battelle Energy Alliance, LLC and is available to a wide variety of university researchers for nuclear fuels and materials experiments as well as medical isotope research and production.

  7. Process kinetics of an activated-sludge reactor system treating poultry slaughterhouse wastewater

    Microsoft Academic Search

    Ting-Hsun Hsiao; Ju-Sheng Huang; Yu-I Huang

    2011-01-01

    The principal objective was to generate the essential kinetic parameters for model simulation and operation management of an activated-sludge reactor (ASR) system treating poultry slaughterhouse wastewater. By varying four different mean cell residence times (?c=4.6–24.3 d), the ASR system (26°C) removed effectively 93.5%–97.2% of chemical oxygen demand (COD) from wastewater. If a high COD removal efficiency and a low effluent volatile

  8. Process kinetics of an activated-sludge reactor system treating poultry slaughterhouse wastewater

    Microsoft Academic Search

    Ting-Hsun Hsiao; Ju-Sheng Huang; Yu-I Huang

    2012-01-01

    The principal objective was to generate the essential kinetic parameters for model simulation and operation management of an activated-sludge reactor (ASR) system treating poultry slaughterhouse wastewater. By varying four different mean cell residence times (?c=4.6–24.3 d), the ASR system (26°C) removed effectively 93.5%–97.2% of chemical oxygen demand (COD) from wastewater. If a high COD removal efficiency and a low effluent volatile

  9. New reactor system for supercritical water oxidation and its application on phenol destruction

    Microsoft Academic Search

    Minok Koo; Won Kook Lee; Choul Ho Lee

    1997-01-01

    A new reactor system for supercritical water oxidation which can treat reaction variables independently was developed. With this system, phenol oxidation experiments were carried out at temperatures 380–440°C and pressures 190–270 atm. Reaction time was varied from 12 to 120 s, and corresponding conversion was 11–99%. The initial phenol concentration was below 8.8 mM based on reaction volume. The initial

  10. Kinetic enhancement of starch bioconversion in thermoseparating aqueous two-phase reactor systems

    Microsoft Academic Search

    Mian Li; Jin-Woo Kim; Tonya L. Peeples

    2002-01-01

    The extractive bioconversion of starch in an aqueous two-phase reactor system (ATPRS) was studied through experimentation and mathematical modeling. The phase-forming components included PEO-PPO-2500 (a random copolymer of ethylene oxide and propylene oxide with molecular weight of 2500) and MgSO4. Partitioning of glucose and maltose in the PEO-PPO\\/MgSO4 system was determined. Hydrolysis rates of soluble and corn starches in one-phase

  11. Neural net controlled tag gas sampling system for nuclear reactors

    DOEpatents

    Gross, K.C.; Laug, M.T.; Lambert, J.B.; Herzog, J.P.

    1997-02-11

    A method and system are disclosed for providing a tag gas identifier to a nuclear fuel rod and analyze escaped tag gas to identify a particular failed nuclear fuel rod. The method and system include disposing a unique tag gas composition into a plenum of a nuclear fuel rod, monitoring gamma ray activity, analyzing gamma ray signals to assess whether a nuclear fuel rod has failed and is emitting tag gas, activating a tag gas sampling and analysis system upon sensing tag gas emission from a failed nuclear rod and evaluating the escaped tag gas to identify the particular failed nuclear fuel rod. 12 figs.

  12. Neural net controlled tag gas sampling system for nuclear reactors

    DOEpatents

    Gross, Kenneth C. (Bolingbrook, IL); Laug, Matthew T. (Idaho Fall, ID); Lambert, John D. B. (Wheaton, IL); Herzog, James P. (Downers Grove, IL)

    1997-01-01

    A method and system for providing a tag gas identifier to a nuclear fuel rod and analyze escaped tag gas to identify a particular failed nuclear fuel rod. The method and system include disposing a unique tag gas composition into a plenum of a nuclear fuel rod, monitoring gamma ray activity, analyzing gamma ray signals to assess whether a nuclear fuel rod has failed and is emitting tag gas, activating a tag gas sampling and analysis system upon sensing tag gas emission from a failed nuclear rod and evaluating the escaped tag gas to identify the particular failed nuclear fuel rod.

  13. Combined on-board hydride slurry storage and reactor system and process for hydrogen-powered vehicles and devices

    DOEpatents

    Brooks, Kriston P; Holladay, Jamelyn D; Simmons, Kevin L; Herling, Darrell R

    2014-11-18

    An on-board hydride storage system and process are described. The system includes a slurry storage system that includes a slurry reactor and a variable concentration slurry. In one preferred configuration, the storage system stores a slurry containing a hydride storage material in a carrier fluid at a first concentration of hydride solids. The slurry reactor receives the slurry containing a second concentration of the hydride storage material and releases hydrogen as a fuel to hydrogen-power devices and vehicles.

  14. Circulation system for flowing uranium hexafluoride cavity reactor experiments

    NASA Technical Reports Server (NTRS)

    Jaminet, J. F.; Kendall, J. S.

    1976-01-01

    Research related to determining the feasibility of producing continuous power from fissile fuel in the gaseous state is presented. The development of three laboratory-scale flow systems for handling gaseous UF6 at temperatures up to 500 K, pressure up to approximately 40 atm, and continuous flow rates up to approximately 50g/s is presented. A UF6 handling system fabricated for static critical tests currently being conducted is described. The system was designed to supply UF6 to a double-walled aluminum core canister assembly at temperatures between 300 K and 400 K and pressure up to 4 atm. A second UF6 handling system designed to provide a circulating flow of up to 50g/s of gaseous UF6 in a closed-loop through a double-walled aluminum core canister with controlled temperature and pressure is described. Data from flow tests using UF6 and UF6/He mixtures with this system at flow rates up to approximately 12g/s and pressure up to 4 atm are presented. A third UF6 handling system fabricated to provide a continuous flow of UF6 at flow rates up to 5g/s and at pressures up to 40 atm for use in rf-heated, uranium plasma confinement experiments is described.

  15. Robotic dismantlement systems at the CP-5 reactor D&D project.

    SciTech Connect

    Seifert, L. S.

    1998-10-28

    The Chicago Pile 5 (CP-5) Research Reactor Facility is currently undergoing decontamination and decommissioning (D&D) at the Argonne National Laboratory (ANL) Illinois site. CP-5 was the principle nuclear reactor used to produce neutrons for scientific research at Argonne from 1954 to 1979. The CP-5 reactor was a heavy-water cooled and moderated, enriched uranium-fueled reactor with a graphite reflector. The CP-5 D&D project includes the disassembly, segmentation and removal of all the radioactive components, equipment and structures associated with the CP-5 facility. The Department of Energy's Robotics Technology Development Program and the Federal Energy Technology Center, Morgantown Office provided teleoperated, remote systems for use in the dismantlement of the CP-5 reactor assembly for tasks requiring remote dismantlement as part of the EM-50 Large-Scale Demonstration Program (LSDP). The teleoperated systems provided were the Dual Arm Work Platform (DAWP), the Rosie Mobile Teleoperated Robot Work System (ROSIE), and a remotely-operated crane control system with installed swing-reduction control system. Another remotely operated apparatus, a Brokk BM250, was loaned to ANL by the Princeton Plasma Physics Laboratory (PPPL). This machine is not teleoperated and was not part of the LSDP, but deserves some mention in this discussion. The DAWP is a robotic dismantlement system that includes a pair of Schilling Robotic Systems Titan III hydraulic manipulator arms mounted to a specially designed support platform: a hydraulic power unit (HPU) and a remote operator console. The DAWP is designed to be crane-suspended for remote positioning. ROSIE, developed by RedZone Robotics, Inc. is a mobile, electro-hydraulic, omnidirectional platform with a heavy-duty telescoping boom mounted to the platform's deck. The work system includes the mobile platform (locomotor), a power distribution unit (PDU) and a remote operator console. ROSIE moves about the reactor building floor around the reactor assembly and, like the DAWP, is controlled from a console in the control room. The remotely-operated crane control system with installed swing-reduction control system was installed on the CP-5 polar crane and allows a load suspended from the crane to be remotely operated while reducing the induced swing in the load, The system includes a remote-controlled rotational hook, two remote-reading load cells and a lightweight portable operator controller. The last component in this discussion, the Brokk BM250, is a commercially-available electro-hydraulically operated demolition tool. A variety of attachments including a 750 lb. jackhammer, hydraulic shear or 1/3 cubic yard bucket can be quickly installed onto its articulated boom. This paper will primarily discuss the teleoperated robotics systems, DAWP and Rosie, their performance, tooling and lessons learned during the dismantlement of the CP-5 reactor structures. Other aspects of the robotics systems' deployment and use such as operator training and maintenance will be briefly discussed as they pertain to the overall performance of the robots.

  16. Gaseous fission product management for molten salt reactors and vented fuel systems

    SciTech Connect

    Messenger, S. J. [Massachusetts Inst. of Technology, 77 Massachusetts Ave., 54-1717, Cambridge, MA 02139 (United States); Forsberg, C. [Massachusetts Inst. of Technology, 77 Massachusetts Ave., 24-207, Cambridge, MA 02139 (United States); Massie, M. [Massachusetts Inst. of Technology, 77 Massachusetts Ave., NW12-230, Cambridge, MA 02139 (United States)

    2012-07-01

    Fission gas disposal is one of the unresolved difficulties for Molten Salt Reactors (MSRs) and advanced reactors with vented fuel systems. As these systems operate, they produce many radioactive isotopes of xenon and krypton (e.g. {sup 135}Xe t{sub 1/2} = 9.14 hours and {sup 85}Kr t{sub 1/2}= 10.73 years). Removing these gases proves vital to the success of such reactor designs for two reasons. First, the gases act as large neutron sinks which decrease reactivity and must be counterbalanced by increasing fuel loading. Second, for MSRs, inert fission product gases naturally separate quickly from high temperature salts, thus creating high vapor pressure which poses safety concerns. For advanced reactors with solid vented fuel, the gases are allowed to escape into an off-gas system and thus must be managed. Because of time delays in transport of fission product gases in vented fuel systems, some of the shorter-lived radionuclides will decay away thereby reducing the fission gas source term relative to an MSR. To calculate the fission gas source term of a typical molten salt reactor, we modeled a 1000 MWe graphite moderated thorium MSR similar to that detailed in Mathieu et al. [1]. The fuel salt used in these calculations was LiF (78 mole percent) - (HN)F 4 (22 mole percent) with a heavy nuclide composition of 3.86% {sup 233}U and 96.14% {sup 232}Th by mass. Before we can remove the fission product gases produced by this reactor configuration, we must first develop an appropriate storage mechanism. The gases could be stored in pressurized containers but then one must be concerned about bottle failure. Methods to trap noble gases in matrices are expensive and complex. Alternatively, there are direct storage/disposal options: direct injection into the Earth or injecting a grout-based product into the Earth. Advances in drilling technologies, hydro fracture technologies, and methods for the sequestration of carbon dioxide from fossil fuel plants are creating new options for disposal of fission gas wastes. In each option, lithostatic pressure, a kilometer or more underground, eliminates the pressure driving force for noble gas release and dissolves any untrapped gas in deep groundwater or into incorporated solid waste forms. The options, challenges, and potential for these methods to dispose of gaseous fission products are described. With this research, we hope to help both MSRs and other advanced reactors come one step closer to commercialization. (authors)

  17. Multi-unit Operations in Non-Nuclear Systems: Lessons Learned for Small Modular Reactors

    SciTech Connect

    OHara J. M.; Higgins, J.; DAgostino, A.

    2012-01-17

    The nuclear-power community has reached the stage of proposing advanced reactor designs to support power generation for decades to come. Small modular reactors (SMRs) are one approach to meet these energy needs. While the power output of individual reactor modules is relatively small, they can be grouped to produce reactor sites with different outputs. Also, they can be designed to generate hydrogen, or to process heat. Many characteristics of SMRs are quite different from those of current plants and may be operated quite differently. One difference is that multiple units may be operated by a single crew (or a single operator) from one control room. The U.S. Nuclear Regulatory Commission (NRC) is examining the human factors engineering (HFE) aspects of SMRs to support licensing reviews. While we reviewed information on SMR designs to obtain information, the designs are not completed and all of the design and operational information is not yet available. Nor is there information on multi-unit operations as envisioned for SMRs available in operating experience. Thus, to gain a better understanding of multi-unit operations we sought the lesson learned from non-nuclear systems that have experience in multi-unit operations, specifically refineries, unmanned aerial vehicles and tele-intensive care units. In this paper we report the lessons learned from these systems and the implications for SMRs.

  18. Treatment of fruit-juice industry wastewater in a two-stage anaerobic hybrid (AH) reactor system followed by a sequencing batch reactor (SBR).

    PubMed

    Tawfik, A; El-Kamah, H

    2012-01-01

    This study has been carried out to assess the performance of a combined system consisting of an anaerobic hybrid (AH) reactor followed by a sequencing batch reactor (SBR) for treatment of fruit-juice industry wastewater at a temperature of 26 degrees C. Three experimental runs were conducted in this investigation. In the first experiment, a single-stage AH reactor was operated at a hydraulic retention time (HRT) of 10.2 h and organic loading rate (OLR) of 11.8 kg COD m(-3) d(-1). The reactor achieved a removal efficiency of 42% for chemical oxygen demand (COD), 50.8% for biochemical oxygen demand (BOD5), 50.3% for volatile fatty acids (VFA) and 56.4% for total suspended solids (TSS). In the second experiment, two AH reactors connected in series achieved a higher removal efficiency for COD (67.4%), BOD5 (77%), and TSS (71.5%) at a total HRT of 20 h and an OLR of 5.9 kg COD m(-3) d(-1). For removal of the remaining portions of COD, BOD5 and TSS from the effluent of the two-stage AH system, a sequencing batch reactor (SBR) was investigated as a post-treatment unit. The reactor achieved a substantial reduction in total COD, resulting in an average effluent concentration of 50 mg L(-1) at an HRT of 11 h and OLR of 5.3 kg COD m(-3) d(-1). Almost complete removal of total BOD5 and oil and grease was achieved, i.e. 10 mg L(-1) and 1.2 mg L(-1), respectively, remained in the final effluent of the SBR. PMID:22629614

  19. 3D Neutronic Analysis in MHD Calculations at ARIES-ST Fusion Reactors Systems

    NASA Astrophysics Data System (ADS)

    Hançerliogullar?, Aybaba; Cini, Mesut

    2013-10-01

    In this study, we developed new models for liquid wall (FW) state at ARIES-ST fusion reactor systems. ARIES-ST is a 1,000 MWe fusion reactor system based on a low aspect ratio ST plasma. In this article, we analyzed the characteristic properties of magnetohydrodynamics (MHD) and heat transfer conditions by using Monte-Carlo simulation methods (ARIES Team et al. in Fusion Eng Des 49-50:689-695, 2000; Tillack et al. in Fusion Eng Des 65:215-261, 2003) . In fusion applications, liquid metals are traditionally considered to be the best working fluids. The working liquid must be a lithium-containing medium in order to provide adequate tritium that the plasma is self-sustained and that the fusion is a renewable energy source. As for Flibe free surface flows, the MHD effects caused by interaction with the mean flow is negligible, while a fairly uniform flow of thick can be maintained throughout the reactor based on 3-D MHD calculations. In this study, neutronic parameters, that is to say, energy multiplication factor radiation, heat flux and fissile fuel breeding were researched for fusion reactor with various thorium and uranium molten salts. Sufficient tritium amount is needed for the reactor to work itself. In the tritium breeding ratio (TBR) >1.05 ARIES-ST fusion model TBR is >1.1 so that tritium self-sufficiency is maintained for DT fusion systems (Starke et al. in Fusion Energ Des 84:1794-1798, 2009; Najmabadi et al. in Fusion Energ Des 80:3-23, 2006).

  20. Space reactor/Stirling cycle systems for high power lunar applications

    NASA Technical Reports Server (NTRS)

    Schmitz, Paul C.; Mason, Lee S.

    1991-01-01

    An analysis is performed to mathematically model a 550 kWe lunar base power supply which uses a SP-100 reactor coupled with Stirling converters. The reactor is placed in an excavation to keep activated coolant in the hole and to allow maintance of the components outside the hole. Two technology levels are considered. They are 1050 and 1300 K heater head Stirling converts. It is found that for a 1050 K converter the total mass which provided 1000 volts dc at 250 m is 14,366 kg while the 1300 K system mass is 12,104 kg. The radiation area of the 1050 and 1300 K systems are 641 and 356 sq m respectively. Comparisons are made with Brayton and thermionic systems with both near term and advanced technology considered.

  1. Space reactor/Stirling cycle systems for high power lunar application

    NASA Technical Reports Server (NTRS)

    Schmitz, Paul C.; Mason, Lee S.

    1991-01-01

    An analysis is performed to mathematically model a 550 kWe lunar base power supply which uses a SP-100 reactor coupled with Stirling converters. The reactor is placed in an excavation to keep activated coolant in the hole and to allow maintenance of the components outside the hole. Two technology levels are considered. They are 1050 and 1300 K heater head Stirling converts. It is found that for a 1050 K converter the total mass which provided 1000 volts DC at 250 m is 14,366 kg while the 1300 K system mass is 12,104 kg. The radiation area of the 1050 and 1300 K systems are 641 and 356 sq m respectively. Comparisons are made with Brayton and thermionic systems with both near term and advanced technology considered.

  2. Engineering and Physics Optimization of Breed and Burn Fast Reactor Systems: Annual and Final Report

    SciTech Connect

    Kevan D. Weaver; Theron Marshall; James Parry

    2005-10-01

    The Idaho National Laboratory (INL) contribution to the Nuclear Energy Research Initiative (NERI) project number 2002-005 was divided into reactor physics, and thermal-hydraulics and plant design. The research targeted credible physics and thermal-hydraulics models for a gas-cooled fast reactor, analyzing various fuel and in-core fuel cycle options to achieve a true breed and burn core, and performing a design basis Loss of Coolant Accident (LOCA) analysis on that design. For the physics analysis, a 1/8 core model was created using different enrichments and simulated equilibrium fuel loadings. The model was used to locate the hot spot of the reactor, and the peak to average energy deposition at that location. The model was also used to create contour plots of the flux and energy deposition over the volume of the reactor. The eigenvalue over time was evaluated using three different fuel configurations with the same core geometry. The breeding capabilities of this configuration were excellent for a 7% U-235 model and good in both a plutonium model and a 14% U-235 model. Changing the fuel composition from the Pu fuel which provided about 78% U-238 for breeding to the 14% U-235 fuel with about 86% U-238 slowed the rate of decrease in the eigenvalue a noticeable amount. Switching to the 7% U-235 fuel with about 93% U-238 showed an increase in the eigenvalue over time. For the thermal-hydraulic analysis, the reactor design used was the one forwarded by the MIT team. This reactor design uses helium coolant, a Brayton cycle, and has a thermal power of 600 MW. The core design parameters were supplied by MIT; however, the other key reactor components that were necessary for a plausible simulation of a LOCA were not defined. The thermal-hydraulic and plant design research concentrated on determining reasonable values for those undefined components. The LOCA simulation was intended to provide insights on the influence of the Reactor Cavity Cooling System (RCCS), the containment building, and a Decay Heat Removal System (DHRS) on the natural circulation heat transfer of the core's decay heat. A baseline case for natural circulation had to be established in order to truly understand the impact of the added safety systems. This baseline case did not include a DHRS, although the current MIT design does have a DHRS that features the highly efficient Printed Circuit Heat Exchangers (PCHEs). The initial LOCA analysis revealed that the RCCS was insufficient to maintain the reactor core below the fuel matrix decomposition temperature. A guard containment was added to the model in order to maintain a prescribed backpressure during the LOCA to enhance the natural circulation. The backpressure approach did provide satisfactory natural convection during the LOCA. The necessary backpressure was 1.8 MPa, which was not especially different from the values reported by other gas fast reactor researchers. However, as the model evolved to be more physically representative of a nuclear reactor, i.e., it included radial peaking factors, inlet plenum orificing, and the degradation of SiC thermal properties as a result of irradiation, the LOCA-induced fuel temperatures were not consistently below the decomposition limit.

  3. DISTRIBUTION SYSTEMS AS RESERVOIRS AND REACTORS FOR INORGANIC CONTAMINANTS

    EPA Science Inventory

    This paper provides a review of numerous drinking water and geochemical investigations and recent studies of pipe deposits and water treatment materials. This analysis shows that there is growing evidence from analogous natural water systems and some analytical studies that many ...

  4. Teleoperated systems for nuclear reactors: Inspection and maintenance

    NASA Technical Reports Server (NTRS)

    Dorokhov, V. P.; Dorokhov, D. V.; Eperin, A. P.

    1994-01-01

    The present paper describes author's work in the field of teleoperated equipment for inspection and maintenance of the RBML technological channels and graphite laying, emergency operations. New technological and design solutions of teleoperated robotic systems developed for Leningradsky Power Plant are discussed.

  5. Direct-Drive Gas-Cooled Reactor Power System: Concept and Preliminary Testing

    NASA Technical Reports Server (NTRS)

    Wright, S. A.; Lipinski, R. J.; Godfroy, T. J.; Bragg-Sitton, S. M.; VanDyke, M. K.

    2002-01-01

    This paper describes the concept and preliminary component testing of a gas-cooled, UN-fueled, pin-type reactor which uses He/Xe gas that goes directly into a recuperated Brayton system to produce electricity for nuclear electric propulsion. This Direct-Drive Gas-Cooled Reactor (DDG) is designed to be subcritical under water or wet- sand immersion in case of a launch accident. Because the gas-cooled reactor can directly drive the Brayton turbomachinery, it is possible to configure the system such that there are no external surfaces or pressure boundaries that are refractory metal, even though the gas delivered to the turbine is 1144 K. The He/Xe gas mixture is a good heat transport medium when flowing, and a good insulator when stagnant. Judicious use of stagnant cavities as insulating regions allows transport of the 1144-K gas while keeping all external surfaces below 900 K. At this temperature super-alloys (Hastelloy or Inconel) can be used instead of refractory metals. Super-alloys reduce the technology risk because they are easier to fabricate than refractory metals, we have a much more extensive knowledge base on their characteristics, and, because they have a greater resistance to oxidation, system testing is eased. The system is also relatively simple in its design: no additional coolant pumps, heat exchanger, or freeze-thaw systems are required. Key to success of this concept is a good knowledge of the heat transfer between the fuel pins and the gas, as well as the pressure drop through the system. This paper describes preliminary testing to obtain this key information, as well as experience in demonstrating electrically heated testing of simulated reactor components.

  6. Development of inspection systems for alloy 600 nozzles of PWR reactor vessel

    SciTech Connect

    Unate, K.; Ideo, M.; Sanagawa, T. [Mitsubishi Heavy Industries, Ltd., Kobe (Japan); Shirai, T.; Araki, Y. [Mitsubishi Heavy Industries, Ltd., Takasago (Japan)

    1995-08-01

    PWR reactor vessels have alloy 600 nozzles at top and bottom heads. The former are head penetration nozzles for CRDM, and the latter are bottom mounted instrumentation nozzles. The authors have developed inspection systems of two types for each nozzle to confirm the soundness. ECT and UT Techniques are employed for both systems. These systems are controlled remotely and enable to reduce radiation exposure, inspection time and number of inspectors. Based on the functional tests using full scale mockups, the reliabilities and effectiveness of both systems were confirmed.

  7. Assessment of compatibility of a system with fast reactors with sustainability requirements and paths to its development

    Microsoft Academic Search

    Baldev Raj; A. Vasile; V. Kagramanian; M. Xu; R. Nakai; Y. I. Kim; V. Usanov; F. Depischg; A. Stanculescu

    The purpose of the paper is to review results of the assessment of an Innovative Nuclear System based on a Closed Nuclear Fuel Cycle with Fast Reactors (INS CNFC-FR) that was performed jointly by Canada, China, France, India, Japan, Republic of Korea, Russia, and Ukraine within the phase I of the IAEA international project for innovative reactors and fuel cycles

  8. Pressure suppression containment system for boiling water reactor

    DOEpatents

    Gluntz, Douglas M. (San Jose, CA); Nesbitt, Loyd B. (San Jose, CA)

    1997-01-01

    A system for suppressing the pressure inside the containment of a BWR following a postulated accident. A piping subsystem is provided which features a main process pipe that communicates the wetwell airspace to a connection point downstream of the guard charcoal bed in an offgas system and upstream of the main bank of delay charcoal beds which give extensive holdup to offgases. The main process pipe is fitted with both inboard and outboard containment isolation valves. Also incorporated in the main process pipe is a low-differential-pressure rupture disk which prevents any gas outflow in this piping whatsoever until or unless rupture occurs by virtue of pressure inside this main process pipe on the wetwell airspace side of the disk exceeding the design opening (rupture) pressure differential. The charcoal holds up the radioactive species in the noncondensable gas from the wetwell plenum by adsorption, allowing time for radioactive decay before the gas is vented to the environs.

  9. Pressure suppression containment system for boiling water reactor

    DOEpatents

    Gluntz, D.M.; Nesbitt, L.B.

    1997-01-21

    A system is disclosed for suppressing the pressure inside the containment of a BWR following a postulated accident. A piping subsystem is provided which features a main process pipe that communicates the wetwell airspace to a connection point downstream of the guard charcoal bed in an offgas system and upstream of the main bank of delay charcoal beds which give extensive holdup to offgases. The main process pipe is fitted with both inboard and outboard containment isolation valves. Also incorporated in the main process pipe is a low-differential-pressure rupture disk which prevents any gas outflow in this piping whatsoever until or unless rupture occurs by virtue of pressure inside this main process pipe on the wetwell airspace side of the disk exceeding the design opening (rupture) pressure differential. The charcoal holds up the radioactive species in the noncondensable gas from the wetwell plenum by adsorption, allowing time for radioactive decay before the gas is vented to the environs. 3 figs.

  10. Space reactor/Stirling cycle systems for high power lunar applications

    SciTech Connect

    Schmitz, P.C. (Sverdrup Technology, Inc., 21000 Brookpark Rd., MS 301-5, Cleveland, Ohio 44135 (US)); Mason, L.S. (NASA Lewis Research Center, 21000 Brookpark Rd., MS 501-6, Cleveland, Ohio 44135 (US))

    1991-01-05

    It is desired to estimate performance and mass of a 550 kWe SP-100/Stirling nuclear power lunar base. Mass and efficiency estimates are made by modeling the components as a function of thermal or electrical power output requirements. It is found that utilizing a 1050 K heater head the total system mass is 13537 kg. For the 1300 K heater head temperature the system mass is 11474 kg. Mass and radiator area comparisons are made with a SP-100/Brayton and an Incore thermionic reactor. Two technology levels are looked at which correspond to low and high temperature systems (for the thermionic system it also includes a increase in thermionic output voltage). Stirling converter systems are the lightest of the low temperatures systems. At higher temperatures all of the systems masses are similar. Thermionic systems always produced the smallest radiators because of their high heat rejection temperature with Stirling systems coming in a close second.

  11. Detailed bifurcation analysis with a simplified model for advance heavy water reactor system

    NASA Astrophysics Data System (ADS)

    Pandey, Vikas; Singh, Suneet

    2015-01-01

    The bifurcation analysis of fixed points and limit cycles with a simplified mathematical model representing system dynamics of a boiling water reactor has been carried out, specifically parameter values for AHWR is used. The lumped parameter model that includes point reactor kinetics equation for neutron balance in the reactor core and one node model for fuel and coolant thermal hydraulics is used in the analysis. The nonlinearity due to reactivity is considered in the present model; while other nonlinearities due to heat transfer process between fuel-clad and fuel-coolant has been neglected. The system loses its stability via Hopf bifurcation as the system parameters are varied. The continuations of subcritical and supercritical Hopf points show the existence of limit point bifurcations of limit cycles (LPC). The codimension one and codimension two bifurcations of fixed points for the system have been analyzed. The stability of observed limit cycles has been analyzed by Floquet multiplier as well as by Lyapunov coefficient. The pattern of limit cycles and envelopes of limit cycles over the fixed points have been studied by numerical integrations and depicted by time history graphs.

  12. Development of an on-line fuel failure monitoring system for CANDU reactors

    NASA Astrophysics Data System (ADS)

    Livingstone, Stephen Jason

    Although relatively rare in CANDU plants, fuel defects have always been an important operating concern for CANDU fuel operation and behaviour, and play a critical role in health, safety, and the economics of an operating reactor. A fuel defect occurs when a fuel element has a breach in its sheath resulting in fission product (FP) release and/or uranium fuel loss into the reactor primary heat transport system (PHTS). The unintended release of FP and fuel material into the PHTS creates elevated radiation fields and hazards for the reactor operator. The intent of this thesis is to develop an online real-time system that can analyse PHTS activity and infer information relating to otherwise unknown defect(s) in the reactor core. A MATLABRTM based Graphical User Interface (GUI) programme called COLDD (CANDU On-Line Defected fuel Diagnostic) has been developed to provide detailed diagnostics of PHTS activities. COLDD is based on new techniques, a new empirical diffusion coefficient, new algorithms, and refinement of existing techniques. Several techniques are based on detailed mechanistic models that are presented in detail, while other techniques are based on empirical rules from experimental and commercial experience; the diversity of techniques are shown to be self-consistent. The techniques employed by COLDD are compared to techniques used internationally by other defected fuel diagnostic tools for non-CANDU type reactors. The ability for COLDD to perform successful defected fuel diagnostics is dependent on the quality of the data provided. A detailed sensitivity analysis is performed to determine key measurement parameters. Techniques are also developed to allow operators to perform robust error checking of data to ensure consistency. COLDD is validated against theoretical, experimental, and commercial reactor data, and shown to be stable and consistent in all cases. The plethora of analysis modes are shown to be self-consistent. The version of COLDD described in this thesis is considered a Beta version ready for testing at a commercial station, and for development to add and improve algorithms and the user interface. There are several possible improvements discussed, including gaps in defected fuel understanding that require further research. COLDD is highly stable and has been demonstrated to be effective; it is a new powerful tool in the arsenal ofa reactor operator faced with defected fuel in core. Key words: Defected fuel, CANDU, Nuclear, Fission Product Release, Diagnostic.

  13. Conceptual Design of HP-STMCs Space Reactor Power System for 110 kWe

    SciTech Connect

    El-Genk, Mohamed S.; Tournier, Jean-Michel [Institute for Space and Nuclear Power Studies, University of New Mexico, Albuquerque, NM, 87131 (United States); Chemical and Nuclear Engineering Dept., University of New Mexico, Albuquerque, NM, 87131 (United States)

    2004-02-04

    A conceptual design of a Heat Pipe-Segmented Thermoelectric Module Converters (HP-STMCs) space reactor power system (SRPS) for a net power of 110 kWe is developed. The parametric analysis changed the number of radiator's potassium heat pipes from 224 to 336 and calculated the effects on the operation parameters and total mass of the system. The reactor has a hexagonal core comprised of 126 heat pipe modules, each consists of three UN, 1.5 cm OD fuel pins brazed to a central lithium heat pipe of identical diameter. The Re cladding of the fuel pins is brazed along the active core length to the lithium heat pipe using 6 Re tri-cusps. The reactor control is accomplished using 12 B4C/BeO control drums, a large diameter one on each side of the hexagonal core and a small diameter one at each corner. The control drums are placed within the radial BeO reflector (7.1-9.1 cm thick). The fuel pin peak-to-average power ratio in the reactor core is 1.12-1.19. Despite its very high density and fabrication challenge, using rhenium structure in the reactor core is necessary for three main reasons: (a) the high reactor temperature ({>=} 1500 K); (b) excellent compatibility with the UN fuel and lithium; (c) to cause a spectrum shift that ensures having sufficient negative reactivity margin during a water submersion accident. The reference HP-STMC system with 324, 2.42-3.03 cm OD potassium heat pipes in the radiator is 9.60 m long and has a cone angle of 30 deg. The nominal operation of the reactor's lithium heat pipes and of the radiator's potassium heat pipes is at or below {approx} 45% of the prevailing wicking and sonic limit, respectively. The masses of the reactor and radiation shadow shield are 753.7 kg and 999.5 kg, respectively; the average heat pipes temperature in the reactor is 1513 K; the mass of the reactor's lithium heat pipes with a C-C finned condenser that is 1.5 m long is 516.1 kg; the mass of the radiator is 557.5 kg, with an outer surface area of 87 m2 (6.41 kg/m2) and effective temperatures of 752 K and 734 K for the front and rear radiator sections, respectively. These estimates are for a constant collector temperature for the STMCs of 1300 K and STMCs' thermal and electrical losses of 5% and 8%, respectively. The estimates of the total mass and specific power of the reference HP-STMCs SRPS, pending future detailed design and analysis, are 4261 kg and 25.8 We/kg, respectively.

  14. Cultivation of shear stress sensitive microorganisms in disposable bag reactor systems.

    PubMed

    Jonczyk, Patrick; Takenberg, Meike; Hartwig, Steffen; Beutel, Sascha; Berger, Ralf G; Scheper, Thomas

    2013-09-20

    Technical scale (?5l) cultivations of shear stress sensitive microorganisms are often difficult to perform, as common bioreactors are usually designed to maximize the oxygen input into the culture medium. This is achieved by mechanical stirrers, causing high shear stress. Examples for shear stress sensitive microorganisms, for which no specific cultivation systems exist, are many anaerobic bacteria and fungi, such as basidiomycetes. In this work a disposable bag bioreactor developed for cultivation of mammalian cells was investigated to evaluate its potential to cultivate shear stress sensitive anaerobic Eubacterium ramulus and shear stress sensitive basidiomycetes Flammulina velutipes and Pleurotus sapidus. All cultivations were compared with conventional stainless steel stirred tank reactors (STR) cultivations. Good growth of all investigated microorganisms cultivated in the bag reactor was found. E. ramulus showed growth rates of ?=0.56 h?¹ (bag) and ?=0.53 h?¹ (STR). Differences concerning morphology, enzymatic activities and growth in fungal cultivations were observed. In the bag reactor growth in form of small, independent pellets was observed while STR cultivations showed intense aggregation. F. velutipes reached higher biomass concentrations (21.2 g l?¹ DCW vs. 16.8 g l?¹ DCW) and up to 2-fold higher peptidolytic activities in comparison to cell cultivation in stirred tank reactors. PMID:23892193

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

    NASA Technical Reports Server (NTRS)

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

    1985-01-01

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

  16. Reduction of Worldwide Plutonium Inventories Using Conventional Reactors and Advanced Fuels: A Systems Study

    SciTech Connect

    Krakowski, R.A., Bathke, C.G.

    1997-12-31

    The potential for reducing plutonium inventories in the civilian nuclear fuel cycle through recycle in LWRs of a variety of mixed oxide forms is examined by means of a cost based plutonium flow systems model. This model emphasizes: (1) the minimization of separated plutonium; (2) the long term reduction of spent fuel plutonium; (3) the optimum utilization of uranium resources; and (4) the reduction of (relative) proliferation risks. This parametric systems study utilizes a globally aggregated, long term (approx. 100 years) nuclear energy model that interprets scenario consequences in terms of material inventories, energy costs, and relative proliferation risks associated with the civilian fuel cycle. The impact of introducing nonfertile fuels (NFF,e.g., plutonium oxide in an oxide matrix that contains no uranium) into conventional (LWR) reactors to reduce net plutonium generation, to increase plutonium burnup, and to reduce exo- reactor plutonium inventories also is examined.

  17. Start-up simulation of a thermionic space nuclear reactor system

    SciTech Connect

    El-Genk, M.S.; Xue, H.; Paramonov, D. (Institute for Space Nuclear Power Studies, Department of Chemical and Nuclear Engineering, The University of New Mexico, Albuquerque, New Mexico 87131-1341 (United States))

    1993-01-15

    The Thermionic Transient Analysis Model (TITAM) is used in this paper to simulate the start-up of the TOPAZ-II space nuclear power system in orbit. The start-up procedures simulated herein are assumed for the purpose of demonstrating the capabilities of the model and may not represent an accurate account of the actual start-up procedures of the TOPAZ-II system. The temperature reactivity feedback effects of the moderator, UO[sub 2] fuel, electrodes, coolant, and other components in the core are calculated and their effects on the thermal and criticality conditions of the reactor are investigated. Also, estimates of the time constants of the temperature reactivity feedback for the UO[sub 2] fuel and the ZrH moderator during start-up, as well as of the total temperature reactivity feedback as a function of the reactor steady-state thermal power, are obtained.

  18. Transient analysis and startup simulation of a thermionic space nuclear reactor system

    SciTech Connect

    El-Genk, M.S.; Xue, Huimin; Paramonov, D. (Univ. of New Mexico, Albuquerque, NM (United States). Dept. of Chemical and Nuclear Engineering)

    1994-01-01

    The thermionic transient analysis model is used to simulate the startup of the TOPAZ-2 space nuclear power system in orbit. The simulated startup procedures are assumed for the purpose of demonstrating the capabilities of the model and may not represent an accurate account of the actual startup procedures of the TOPAZ-2 system. The temperature reactivity feedback effects of the moderator, UO[sub 2] fuel, electrodes, coolant, and other components in the core are calculated, and their effects on the thermal and criticality conditions of the reactor are investigated. Also, estimates of the time constants of the temperature reactivity feedback for the UO[sub 2] fuel and the ZrH moderator during startup, as well as of the total temperature reactivity feedback as a function of the reactor steady-state thermal power, are obtained.

  19. Analysis of Three Mile Island Unit 2 reactor cooling system transients. Volume 5

    SciTech Connect

    Henrie, J.O.; Postma, A.K.

    1983-08-01

    The reactimeter data recorded on 3/28/79 at Three Mile Island (TMI-2) indicates a number of abrupt transients starting at 13:52. These transients appeared to be the results of very rapid energy releases in the reactor cooling system. A study was initiated by the US Department of Energy to determine the causes and consequences of these transients. The study shows that the transients were not caused by energy releases in the reactor cooling system. They were probably caused by malfunctions in the reactimeter power supply or by reactimeter ground loop faults. Information obtained and observations derived from the study of real energy release transients which occurred during the first day of the TMI-2 loss of coolant accident are presented.

  20. Scaling Analysis for the Direct Reactor Auxiliary Cooling System for AHTRs

    SciTech Connect

    Yoder Jr, Graydon L [ORNL] [ORNL; Wilson, Dane F [ORNL] [ORNL; Wang, X. NMN [Ohio State University] [Ohio State University; Lv, Q. NMN [Ohio State University] [Ohio State University; Sun, X NMN [Ohio State University] [Ohio State University; Christensen, R. N. [Ohio State University] [Ohio State University; Blue, T. E. [Ohio State University] [Ohio State University; Subharwall, Piyush [Idaho National Laboratory (INL)] [Idaho National Laboratory (INL)

    2011-01-01

    The Direct Reactor Auxiliary Cooling System (DRACS), shown in Fig. 1 [1], is a passive heat removal system proposed for the Advanced High-Temperature Reactor (AHTR). It features three coupled natural circulation/convection loops completely relying on the buoyancy as the driving force. A prototypic design of the DRACS employed in a 20-MWth AHTR has been discussed in our previous work [2]. The total height of the DRACS is usually more than 10 m, and the required heating power will be large (on the order of 200 kW), both of which make a full-scale experiment not feasible in our laboratory. This therefore motivates us to perform a scaling analysis for the DRACS to obtain a scaled-down model. In this paper, theory and methodology for such a scaling analysis are presented.

  1. System and method for determining coolant level and flow velocity in a nuclear reactor

    DOEpatents

    Brisson, Bruce William; Morris, William Guy; Zheng, Danian; Monk, David James; Fang, Biao; Surman, Cheryl Margaret; Anderson, David Deloyd

    2013-09-10

    A boiling water reactor includes a reactor pressure vessel having a feedwater inlet for the introduction of recycled steam condensate and/or makeup coolant into the vessel, and a steam outlet for the discharge of produced steam for appropriate work. A fuel core is located within a lower area of the pressure vessel. The fuel core is surrounded by a core shroud spaced inward from the wall of the pressure vessel to provide an annular downcomer forming a coolant flow path between the vessel wall and the core shroud. A probe system that includes a combination of conductivity/resistivity probes and/or one or more time-domain reflectometer (TDR) probes is at least partially located within the downcomer. The probe system measures the coolant level and flow velocity within the downcomer.

  2. Modular hybrid plasma reactor and related systems and methods

    DOEpatents

    Kong, Peter C.; Grandy, Jon D.; Detering, Brent A.

    2010-06-22

    A device, method and system for generating a plasma is disclosed wherein an electrical arc is established and the movement of the electrical arc is selectively controlled. In one example, modular units are coupled to one another to collectively define a chamber. Each modular unit may include an electrode and a cathode spaced apart and configured to generate an arc therebetween. A device, such as a magnetic or electromagnetic device, may be used to selectively control the movement of the arc about a longitudinal axis of the chamber. The arcs of individual modules may be individually controlled so as to exhibit similar or dissimilar motions about the longitudinal axis of the chamber. In another embodiment, an inlet structure may be used to selectively define the flow path of matter introduced into the chamber such that it travels in a substantially circular or helical path within the chamber.

  3. Hydraulic Shuttle Irradiation System (HSIS) Recently Installed in the Advanced Test Reactor (ATR)

    Microsoft Academic Search

    A. Joseph Palmer; Gerry L. McCormick; Shannon J. Corrigan

    2010-01-01

    2010 International Congress on Advances in Nuclear Power Plants (ICAPP10) ANS Annual Meeting Imbedded Topical San Diego, CA June 13 17, 2010 Hydraulic Shuttle Irradiation System (HSIS) Recently Installed in the Advanced Test Reactor (ATR) Author: A. Joseph Palmer, Mechanical Engineer, Irradiation Test Programs, 208-526-8700, Joe.Palmer@INL.gov Affiliation: Idaho National Laboratory P.O. Box 1625, MS-3840 Idaho Falls, ID 83415 INL\\/CON-10-17680 ABSTRACT

  4. Safety system consideration of a supercritical-water cooled fast reactor with simplified PSA

    Microsoft Academic Search

    J. H Lee; Y Oka; S Koshizuka

    1999-01-01

    The probabilistic safety of the supercritical-water cooled fast reactor (SCFR) is evaluated with the simplified probabilistic safety assessment (PSA) methodology. SCFR has a once-through direct cycle where all feedwater flows through the core to the turbine at supercritical pressure. There are no recirculation loops in the once-through direct cycle system, which is the most important difference from the current light

  5. Current practice and developmental efforts for leak detection in US reactor primary systems

    Microsoft Academic Search

    D. S. Kupperman; T. N. Claytor

    1985-01-01

    Current leak detection practices in 74 operating nuclear reactors have been reviewed. Existing leak detection systems are adequate to ensure a leak-before-break scenario in most situations, but no currently available, single method combines optimal leakage detection sensitivity, leak-locating ability, and leakage measurement accuracy. Simply tightening current leakage limits may produce an unacceptably large number of unnecessary shutdowns. The use of

  6. Performance Analysis of Potassium Heat Pipes Radiator for HP-STMCs Space Reactor Power System

    Microsoft Academic Search

    Mohamed S. El-Genk; Jean-Michel Tournier

    2004-01-01

    A detailed design and performance results of C-C finned, and armored potassium heat pipes radiator for a 110 kWe Heat Pipes-Segmented Thermoelectric Module Converters (HP-STMCs) Space Reactor Power system (SRPS) are presented. The radiator consists of two sections; each serves an equal number of STMCs and has 162 longitudinal potassium heat pipes with 0.508 mm thick C-C fins. The width

  7. Nuclear propulsion systems for orbit transfer based on the particle bed reactor

    Microsoft Academic Search

    J. R. Powell; H. Ludewig; F. L. Horn; K. Araj; R. Benenati; O. Lazareth; G. Slovik; M. Solon; W. Tappe; J. Belisle

    1987-01-01

    The technology of nuclear direct propulsion orbit transfer systems based on the Particle Bed Reactor (PBR) is described. A 200 megawatt illustrative design is presented for LEO to GEO and other high ..delta..V missions. The PBR-NOTV can be used in a one-way mode with the shuttle or an expendable launch vehicle, e.g., the Titan 34D7, or as a two-way reusable

  8. AUTOLOAD, an automatic optimal pressurized water reactor reload design system with an expert module

    Microsoft Academic Search

    Z. Li; S. H. Levine

    1994-01-01

    An automatic optimal pressurized water reactor (PWR) reload design expert system AUTOLOAD has been developed. It employs two important new techniques. The first is a new loading priority scheme that defines the optimal placement of the fuel in the core that has the maximum end-of-cycle state k[sub eff]. The second is a new power-shape-driven progressive iteration method for automatically determining

  9. Development of a current 'crowbar' switching system for a nuclear fusion reactor

    Microsoft Academic Search

    J. S. Chang; Y. Ichikawa; T. Kaneda

    1981-01-01

    One engineering aspect of fusion reactors that is still in the primitive stage is the development of a crowbar method for short-circuiting the high-power L-C circuits. At McMaster University the study and development of a crowbar method using a shock-heated molecular or atomic beam plasma is in progress. The novel feature of this system is the addition of a second

  10. Calculation of the dynamic response of reactor containment systems to full core explosions

    Microsoft Academic Search

    M. S. Cowler; N. E. Hoskin; A. G. Rowlinson

    1975-01-01

    ASTARTE, a comprehensive time-dependent two-dimensional Lagrangian hydrodynamic code has been developed to determine numerically the loadings and strains arising within the primary containment system following a high-energy reactor excursion. The detail, both temporal and spatial, which is achievable by judicious use of such a code allows the safety analyst to examine the response of individual sections of the containment. It

  11. Thermionic reactor ion propulsion system /TRIPS/ - Its multi-mission capability.

    NASA Technical Reports Server (NTRS)

    Peelgren, M. L.

    1972-01-01

    The unmanned planetary exploration to be conducted in the last two decades of this century includes many higher energy missions which tax all presently available propulsion systems beyond their limit. One candidate with the versatility and performance to meet these mission objectives is nuclear electric propulsion (NEP). Additionally, the NEP System is feasible in orbit raising operations with the Shuttle or Shuttle/Tug combination. A representative planetary mission is described (Uranus-Neptune flyby with probe), and geocentric performance and tradeoffs are discussed. The NEP System is described in more detail with particular emphasis on the power subsystem consisting of the thermionic reactor, heat rejection subsystem, and neutron shield.

  12. BOLD VENTURE COMPUTATION SYSTEM for nuclear reactor core analysis, Version III

    SciTech Connect

    Vondy, D.R.; Fowler, T.B.; Cunningham, G.W. III.

    1981-06-01

    This report is a condensed documentation for VERSION III of the BOLD VENTURE COMPUTATION SYSTEM for nuclear reactor core analysis. An experienced analyst should be able to use this system routinely for solving problems by referring to this document. Individual reports must be referenced for details. This report covers basic input instructions and describes recent extensions to the modules as well as to the interface data file specifications. Some application considerations are discussed and an elaborate sample problem is used as an instruction aid. Instructions for creating the system on IBM computers are also given.

  13. Comparison of KENO-VI and MCNP5 Criticality Analyses for a Lunar Regolith Clustered-Reactor System

    SciTech Connect

    Bess, John Darrell [Center for Space Nuclear Research PO Box 1625, MS 5855 Idaho Falls, ID 83415-3855 (United States)

    2008-01-21

    The Lunar Regolith Clustered-Reactor System design has been presented as an alternative method for providing surface power to a lunar facility using a fast-fission, heatpipe-cooled nuclear reactor. The reactor system is divided into subcritical units that can be safely launched into orbit without risk of inadvertent criticality in the event of a launch accident. The reactor subunits are emplaced into the lunar surface to form a clustered-reactor system, utilizing the regolith as both radiation shielding and neutron-reflector material. Coordinated placement of multiple subunits can provision a critical reactor system proportional to localized lunar surface power demand. Reactor units assembled using proven and tested materials in radiation environments such as UO{sub 2} fuel, stainless-steel cladding and support, and compatible liquid-metal heatpipes promote safety and reliability, with ease of manufacture and testing. Reactor power levels of approximately 100 kWth per subunit significantly reduces the negative effects of elevated temperature and radiation environments associated with single nuclear power reactors operated at higher power levels. The analysis of subunit criticality in various accident scenarios differs by up to 4% ({approx}$6 in reactivity) between results generated using conventional criticality analysis codes, MCNP5 and KENO-VI. A demonstrated trend exists between results of the two criticality codes as accident conditions approach a multiplication factor of one. Code comparison of a tri-cluster system on the lunar surface provides comparable results with calculated system reactivity within 0.5%. Iron concentration is confirmed as the dominant element in the lunar regolith influencing system reactivity.

  14. Safety System Design Concept and Performance Evaluation for a Long Operating Cycle Simplified Boiling Water Reactor

    Microsoft Academic Search

    Kenji Arai; Seijiro Suzuki; Mikihide Nakamaru; Hideaki Heki

    2003-01-01

    The long operating cycle simplified boiling water reactor is a reactor concept that pursues both safety and the economy by employing a natural circulation reactor core without a refueling, a passive decay heat removal, and an integrated building for the reactor and turbine. Throughout the entire spectrum of the design basis accident, the reactor core is kept covered by the

  15. R and D of On-line Reprocessing Technology for Molten-Salt Reactor Systems

    Microsoft Academic Search

    Jan Uhlir; Radka Tulackova; Karolina Chuchvalcova Bimova

    2006-01-01

    The Molten Salt Reactor (MSR) represents one of promising future nuclear reactor concept included in the Generation IV reactors family. The reactor can be operated as the thorium breeder or as the actinide transmuter. However, the future deployment of Molten-Salt Reactors will be significantly dependent on the successful mastering of advanced reprocessing technologies dedicated to their fuel cycle. Here the

  16. CFD Model Development and validation for High Temperature Gas Cooled Reactor Cavity Cooling System (RCCS) Applications

    SciTech Connect

    Hassan, Yassin; Corradini, Michael; Tokuhiro, Akira; Wei, Thomas Y.C.

    2014-07-14

    The Reactor Cavity Cooling Systems (RCCS) is a passive safety system that will be incorporated in the VTHR design. The system was designed to remove the heat from the reactor cavity and maintain the temperature of structures and concrete walls under desired limits during normal operation (steady-state) and accident scenarios. A small scale (1:23) water-cooled experimental facility was scaled, designed, and constructed in order to study the complex thermohydraulic phenomena taking place in the RCCS during stead-state and transient conditions. The facility represents a portion of the reactor vessel with nine stainless steel coolant risers and utilizes water as coolant. The facility was equipped with instrumentation to measure temperatures and flow rates and a general verification was completed during the shakedown. A model of the experimental facility was prepared using RELAP5-3D and simulations were performed to validate the scaling procedure. The experimental data produced during the stead-state run were compared with the simulation results obtained using RELAP5-3D. The overall behavior of the facility met the expectations. The facility capabilities were confirmed to be very promising in performing additional experimental tests, including flow visualization, and produce data for code validation.

  17. Mechanical design and construction new transport reactor system. Second quarterly progress report, January--March 1995

    SciTech Connect

    NONE

    1995-04-01

    During the last quarter, the detailed mechanical design of the new reactor system was completed and construction of the unit was well underway. The new design includes a mixing zone, riser reactor, cyclone, and downcomer as well as instrumentation, heating elements, insulation, and a structural system for supporting the unit. Design modifications were also made to the hydrocarbon feed system. There were no changes required for the downstream sections which cool and condition the reactor product gas, recover liquid products (if any), and measure product gas make. Construction of the unit is expected to be completed by early May, with shakedown runs beginning immediately after. Installation of the electrical windings, insulation of the unit, erection, hook-up, and checkout are the main items yet to be completed. It is expected that the unit will be ready for test work in the latter part of May. The initial tests planned are both pyrolysis runs and partial oxidation runs using a simulated aromatic naphtha feed. Later this year, heavier hydrocarbon feeds will be tested.

  18. Advanced Fusion Reactors for Space Propulsion and Power Systems

    SciTech Connect

    Chapman, John J.

    2011-06-15

    In recent years the methodology proposed for conversion of light elements into energy via fusion has made steady progress. Scientific studies and engineering efforts in advanced fusion systems designs have introduced some new concepts with unique aspects including consideration of Aneutronic fuels. The plant parameters for harnessing aneutronic fusion appear more exigent than those required for the conventional fusion fuel cycle. However aneutronic fusion propulsion plants for Space deployment will ultimately offer the possibility of enhanced performance from nuclear gain as compared to existing ionic engines as well as providing a clean solution to Planetary Protection considerations and requirements. Proton triggered 11Boron fuel (p- 11B) will produce abundant ion kinetic energy for In-Space vectored thrust. Thus energetic alpha particles' exhaust momentum can be used directly to produce high Isp thrust and also offer possibility of power conversion into electricity. p-11B is an advanced fusion plant fuel with well understood reaction kinematics but will require some new conceptual thinking as to the most effective implementation.

  19. Advanced Fusion Reactors for Space Propulsion and Power Systems

    NASA Technical Reports Server (NTRS)

    Chapman, John J.

    2011-01-01

    In recent years the methodology proposed for conversion of light elements into energy via fusion has made steady progress. Scientific studies and engineering efforts in advanced fusion systems designs have introduced some new concepts with unique aspects including consideration of Aneutronic fuels. The plant parameters for harnessing aneutronic fusion appear more exigent than those required for the conventional fusion fuel cycle. However aneutronic fusion propulsion plants for Space deployment will ultimately offer the possibility of enhanced performance from nuclear gain as compared to existing ionic engines as well as providing a clean solution to Planetary Protection considerations and requirements. Proton triggered 11Boron fuel (p- 11B) will produce abundant ion kinetic energy for In-Space vectored thrust. Thus energetic alpha particles "exhaust" momentum can be used directly to produce high ISP thrust and also offer possibility of power conversion into electricity. p- 11B is an advanced fusion plant fuel with well understood reaction kinematics but will require some new conceptual thinking as to the most effective implementation.

  20. A system analysis computer model for the High Flux Isotope Reactor (HFIRSYS Version 1)

    SciTech Connect

    Sozer, M.C.

    1992-04-01

    A system transient analysis computer model (HFIRSYS) has been developed for analysis of small break loss of coolant accidents (LOCA) and operational transients. The computer model is based on the Advanced Continuous Simulation Language (ACSL) that produces the FORTRAN code automatically and that provides integration routines such as the Gear`s stiff algorithm as well as enabling users with numerous practical tools for generating Eigen values, and providing debug outputs and graphics capabilities, etc. The HFIRSYS computer code is structured in the form of the Modular Modeling System (MMS) code. Component modules from MMS and in-house developed modules were both used to configure HFIRSYS. A description of the High Flux Isotope Reactor, theoretical bases for the modeled components of the system, and the verification and validation efforts are reported. The computer model performs satisfactorily including cases in which effects of structural elasticity on the system pressure is significant; however, its capabilities are limited to single phase flow. Because of the modular structure, the new component models from the Modular Modeling System can easily be added to HFIRSYS for analyzing their effects on system`s behavior. The computer model is a versatile tool for studying various system transients. The intent of this report is not to be a users manual, but to provide theoretical bases and basic information about the computer model and the reactor.

  1. Identification and characterization of passive safety system and inherent safety feature building blocks for advanced light-water reactors

    SciTech Connect

    Forsberg, C.W.

    1989-01-01

    Oak Ridge National Laboratory (ORNL) is investigating passive and inherent safety options for Advanced Light-Water Reactors (ALWRs). A major activity in 1989 includes identification and characterization of passive safety system and inherent safety feature building blocks, both existing and proposed, for ALWRs. Preliminary results of this work are reported herein. This activity is part of a larger effort by the US Department of Energy, reactor vendors, utilities, and others in the United States to develop improved LWRs. The Advanced Boiling Water Reactor (ABWR) program and the Advanced Pressurized Water Reactor (APWR) program have as goals improved, commercially available LWRs in the early 1990s. The Advanced Simplified Boiling Water Reactor (ASBWR) program and the AP-600 program are developing more advanced reactors with increased use of passive safety systems. It is planned that these reactors will become commercially available in the mid 1990s. The ORNL program is an exploratory research program for LWRs beyond the year 2000. Desired long-term goals for such reactors include: (1) use of only passive and inherent safety, (2) foolproof against operator errors, (3) malevolence resistance against internal sabotage and external assault and (4) walkaway safety. The acronym ''PRIME'' (Passive safety, Resilient operation, Inherent safety, Malevolence resistance, and Extended (walkaway) safety) is used to summarize these desired characteristics. Existing passive and inherent safety options are discussed in this document.

  2. Reliability analysis of a passive cooling system using a response surface with an application to the Flexible Conversion Ratio Reactor

    E-print Network

    Fong, Christopher J. (Christopher Joseph)

    2008-01-01

    A comprehensive risk-informed methodology for passive safety system design and performance assessment is presented and demonstrated on the Flexible Conversion Ratio Reactor (FCRR). First, the methodology provides a framework ...

  3. Coproduction of hydrogen and methane via anaerobic fermentation of cornstalk waste in continuous stirred tank reactor integrated with up-flow anaerobic sludge bed.

    PubMed

    Cheng, Xi-Yu; Li, Qian; Liu, Chun-Zhao

    2012-06-01

    A 10 L continuous stirred tank reactor (CSTR) system was developed for a two-stage hydrogen fermentation process with an integrated alkaline treatment. The maximum hydrogen production rate reached 218.5 mL/L h at a cornstalk concentration of 30 g/L, and the total hydrogen yield and volumetric hydrogen production rate reached 58.0 mL/g-cornstalk and 0.55-0.57 L/L d, respectively. A 10 L up-flow anaerobic sludge bed (UASB) was used for continuous methane fermentation of the effluents obtained from the two-stage hydrogen fermentation. At the optimal organic loading rate of 15.0 g-COD/Ld, the COD removal efficiency and volumetric biogas production rate reached 83.3% and 4.6L/Ld, respectively. Total methane yield reached 200.9 mL/g-cornstalk in anaerobic fermentation with the effluents and alkaline hydrolysate. As a result, the total energy recovery by coproduction of hydrogen and methane with anaerobic fermentation of cornstalk reached 67.1%. PMID:22487130

  4. A Project Management and Systems Engineering Structure for a Generation IV Very High Temperature Reactor

    SciTech Connect

    Ed Gorski; Dennis Harrell; Finis Southworth

    2004-09-01

    The Very High Temperature Reactor (VHTR) will be an advanced, very high temperature (approximately 1000o C. coolant outlet temperature), gas cooled nuclear reactor and is the nearest term of six Generation IV reactor technologies for nuclear assisted hydrogen production. In 2001, the Generation IV International Forum (GIF), a ten nation international forum working together with the Department of Energy’s (DOE) Nuclear Energy Research Advisory Committee (NERAC), agreed to proceed with the development of a technology roadmap and identified the next generation of nuclear reactor systems for producing new sources of power. Since a new reactor has not been licensed in the United States since the 1970s, the risks are too large for a single utility to assume in the development of an unprecedented Generation IV reactor. The government must sponsor and invest in the research to resolve major first of a kind (FOAK) issues through a full-scale demonstration prior to industry implementation. DOE’s primary mission for the VHTR is to demonstrate nuclear reactor assisted cogeneration of electricity and hydrogen while meeting the Generation IV goals for safety, sustainability, proliferation resistance and physical security and economics. The successful deployment of the VHTR as a demonstration project will aid in restarting the now atrophied U.S. nuclear power industry infrastructure. It is envisioned that VHTR project participants will include DOE Laboratories, industry partners such as designers, constructors, manufacturers, utilities, and Generation IV international countries. To effectively mange R&D, engineering, procurement, construction, and operation for this multi-organizational and technologically complex project, systems engineering will be used extensively to ensure delivery of the final product. Although the VHTR is an unprecedented FOAK system, the R&D, when assessed using the Office of Science and Technology Gate Model, falls primarily in the 3rd - Exploratory Development, 4th – Advanced Development, and 5th- Engineering Development stages of maturity rather than in the basic and viability stages. Therefore the R&D must be controlled and project driven from the top down to address specific issues of feasibility, proof of design or support of engineering. The design evolution must be through the systems approach including an iterative process of high-level requirements definition, engineering to focus R&D to verify feasibility, requirements development and conceptual design, R&D to verify design and refine detailed requirements for final detailed design. This paper will define a framework for project management and application of systems engineering at the Idaho National Engineering and Environmental Laboratory (INEEL). The VHTR Project includes an overall reactor design and construction activity and four major supporting activities: fuel development and qualification, materials selection and qualification, NRC licensing and regulatory support, and the hydrogen production plant.

  5. Dynamic Modeling and Control of Nuclear Reactors Coupled to Closed-Loop Brayton Cycle Systems using SIMULINK{sup TM}

    SciTech Connect

    Wright, Steven A.; Sanchez, Travis [Sandia National Laboratories, Org 6872 MS-1146, PO Box 5800 Albuquerque, New Mexico 87185 (United States)

    2005-02-06

    The operation of space reactors for both in-space and planetary operations will require unprecedented levels of autonomy and control. Development of these autonomous control systems will require dynamic system models, effective control methodologies, and autonomous control logic. This paper briefly describes the results of reactor, power-conversion, and control models that are implemented in SIMULINK{sup TM} (Simulink, 2004). SIMULINK{sup TM} is a development environment packaged with MatLab{sup TM} (MatLab, 2004) that allows the creation of dynamic state flow models. Simulation modules for liquid metal, gas cooled reactors, and electrically heated systems have been developed, as have modules for dynamic power-conversion components such as, ducting, heat exchangers, turbines, compressors, permanent magnet alternators, and load resistors. Various control modules for the reactor and the power-conversion shaft speed have also been developed and simulated. The modules are compiled into libraries and can be easily connected in different ways to explore the operational space of a number of potential reactor, power-conversion system configurations, and control approaches. The modularity and variability of these SIMULINK{sup TM} models provides a way to simulate a variety of complete power generation systems. To date, both Liquid Metal Reactors (LMR), Gas Cooled Reactors (GCR), and electric heaters that are coupled to gas-dynamics systems and thermoelectric systems have been simulated and are used to understand the behavior of these systems. Current efforts are focused on improving the fidelity of the existing SIMULINK{sup TM} modules, extending them to include isotopic heaters, heat pipes, Stirling engines, and on developing state flow logic to provide intelligent autonomy. The simulation code is called RPC-SIM (Reactor Power and Control-Simulator)

  6. Dynamic Modeling and Control of Nuclear Reactors Coupled to Closed-Loop Brayton Cycle Systems using SIMULINK™

    NASA Astrophysics Data System (ADS)

    Wright, Steven A.; Sanchez, Travis

    2005-02-01

    The operation of space reactors for both in-space and planetary operations will require unprecedented levels of autonomy and control. Development of these autonomous control systems will require dynamic system models, effective control methodologies, and autonomous control logic. This paper briefly describes the results of reactor, power-conversion, and control models that are implemented in SIMULINK™ (Simulink, 2004). SIMULINK™ is a development environment packaged with MatLab™ (MatLab, 2004) that allows the creation of dynamic state flow models. Simulation modules for liquid metal, gas cooled reactors, and electrically heated systems have been developed, as have modules for dynamic power-conversion components such as, ducting, heat exchangers, turbines, compressors, permanent magnet alternators, and load resistors. Various control modules for the reactor and the power-conversion shaft speed have also been developed and simulated. The modules are compiled into libraries and can be easily connected in different ways to explore the operational space of a number of potential reactor, power-conversion system configurations, and control approaches. The modularity and variability of these SIMULINK™ models provides a way to simulate a variety of complete power generation systems. To date, both Liquid Metal Reactors (LMR), Gas Cooled Reactors (GCR), and electric heaters that are coupled to gas-dynamics systems and thermoelectric systems have been simulated and are used to understand the behavior of these systems. Current efforts are focused on improving the fidelity of the existing SIMULINK™ modules, extending them to include isotopic heaters, heat pipes, Stirling engines, and on developing state flow logic to provide intelligent autonomy. The simulation code is called RPC-SIM (Reactor Power and Control-Simulator).

  7. Whole-Pin Furnace system: An experimental facility for studying irradiated fuel pin behavior under potential reactor accident conditions

    SciTech Connect

    Liu, Y.Y.; Tsai, H.C.; Donahue, D.A.; Pushis, D.O.; Savoie, F.E.; Holland, J.W.; Wright, A.E.; August, C.; Bailey, J.L.; Patterson, D.R.

    1990-05-01

    The whole-pin furnace system is a new in-cell experimental facility constructed to investigate how irradiated fuel pins may fail under potential reactor accident conditions. Extensive checkouts have demonstrated excellent performance in remote operation, temperature control, pin breach detection, and fission gas handling. The system is currently being used in testing of EBIR-II-irradiated Integral Fast Reactor (IFR) metal fuel pins; future testing will include EBR-II-irradiated mixed-oxide fuel pins. 7 refs., 4 figs.

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

    NASA Technical Reports Server (NTRS)

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

    1973-01-01

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

  9. Compact, high-power nuclear reactor systems based on small diameter particulate fuel

    Microsoft Academic Search

    J. R. Powell; T. E. Botts

    1982-01-01

    Two compact, high-power nuclear reactor concepts are discussed. Both are gas-cooled cavity-type reactors which utilize particulate fuel of the type now used in HTGR reactors. Unshielded reactor volumes are on the order of one cubic meter. The Fixed Bed Reactor operating temperature is limited to 2500 K and the output power to 250 MW(e). In the Rotating Bed Reactor fuel

  10. Reactor Materials Program -- weldment component toughness of SRS PWS piping materials. [Process Water System

    SciTech Connect

    Sindelar, R.L.

    1993-02-01

    The mechanical properties of austenitic stainless steel materials from the reactor systems in the unirradiated (baseline) and the irradiated conditions have been developed previously for structural and fracture analyses of the pressure boundary of the SRS reactor Process Water System (PWS) components. Individual mechanical specimen test results were compiled into three separate weldment components or regions, namely, the base, weld, and weld heat-affected-zone (HAZ), for two orientations (L-C and C-L) with respect to the pipe axis of the source materials and for two test temperatures of 25 and 125[degrees]C. Twelve separate categories were thus defined to assess the effect of test conditions on the mechanical properties and to facilitate selection of properties for structural and fracture analyses. The testing results show high fracture toughness of the materials and support the demonstration of PWS pressure boundary structural integrity under all conditions of reactor operation. The fracture toughness of a fourth weldment component, namely, the weld fusion line region, has been measured to evaluate the potential for a region of low toughness in the interface between the Type 308 stainless steel weld metal and the Type 304 stainless steel pipe. The testing details and results of the weld fusion line toughness are contained in this report.

  11. Aging assessment of the boiling-water reactor (BWR) standby liquid control system

    SciTech Connect

    Orton, R.D.; Johnson, A.B.; Buckley, G.D.; Larson, L.L.

    1992-10-01

    Pacific Northwest Laboratory conducted a Phase I aging assessment of the standby liquid control (SLC) system used in boiling-water reactors. The study was based on detailed reviews of SLC system component and operating experience information obtained from the Nuclear Plant Reliability Database System, the Nuclear Document System, Licensee Event Reports, and other databases. Sources dealing with sodium pentaborate, borates, boric acid, and the effects of environment and corrosion in the SLC system were reviewed to characterize chemical properties and corrosion characteristics of borated solutions. The leading aging degradation concern to date appears to be setpoint drift in relief valves, which has been discovered during routine surveillance and is thought to be caused by mechanical wear. Degradation was also observed in pump seals and internal valves. In general, however, the results of the Phase I study suggest that age-related degradation of SLC systems has not been serious.

  12. Aging assessment of the boiling-water reactor (BWR) standby liquid control system. Phase 1

    SciTech Connect

    Orton, R.D.; Johnson, A.B.; Buckley, G.D.; Larson, L.L.

    1992-10-01

    Pacific Northwest Laboratory conducted a Phase I aging assessment of the standby liquid control (SLC) system used in boiling-water reactors. The study was based on detailed reviews of SLC system component and operating experience information obtained from the Nuclear Plant Reliability Database System, the Nuclear Document System, Licensee Event Reports, and other databases. Sources dealing with sodium pentaborate, borates, boric acid, and the effects of environment and corrosion in the SLC system were reviewed to characterize chemical properties and corrosion characteristics of borated solutions. The leading aging degradation concern to date appears to be setpoint drift in relief valves, which has been discovered during routine surveillance and is thought to be caused by mechanical wear. Degradation was also observed in pump seals and internal valves. In general, however, the results of the Phase I study suggest that age-related degradation of SLC systems has not been serious.

  13. Facile synthesis of graphene on dielectric surfaces using a two-temperature reactor CVD system.

    PubMed

    Zhang, C; Man, B Y; Yang, C; Jiang, S Z; Liu, M; Chen, C S; Xu, S C; Sun, Z C; Gao, X G; Chen, X J

    2013-10-01

    Direct deposition of graphene on a dielectric substrate is demonstrated using a chemical vapor deposition system with a two-temperature reactor. The two-temperature reactor is utilized to offer sufficient, well-proportioned floating Cu atoms and to provide a temperature gradient for facile synthesis of graphene on dielectric surfaces. The evaporated Cu atoms catalyze the reaction in the presented method. C atoms and Cu atoms respectively act as the nuclei for forming graphene film in the low-temperature zone and the zones close to the high-temperature zones. A uniform and high-quality graphene film is formed in an atmosphere of sufficient and well-proportioned floating Cu atoms. Raman spectroscopy, scanning electron microscopy and atomic force microscopy confirm the presence of uniform and high-quality graphene. PMID:24013529

  14. Expert system for identification of simultaneous and sequential reactor fuel failures with gas tagging

    DOEpatents

    Gross, K.C.

    1994-07-26

    Failure of a fuel element in a nuclear reactor core is determined by a gas tagging failure detection system and method. Failures are catalogued and characterized after the event so that samples of the reactor's cover gas are taken at regular intervals and analyzed by mass spectroscopy. Employing a first set of systematic heuristic rules which are applied in a transformed node space allows the number of node combinations which must be processed within a barycentric algorithm to be substantially reduced. A second set of heuristic rules treats the tag nodes of the most recent one or two leakers as background'' gases, further reducing the number of trial node combinations. Lastly, a fuzzy'' set theory formalism minimizes experimental uncertainties in the identification of the most likely volumes of tag gases. This approach allows for the identification of virtually any number of sequential leaks and up to five simultaneous gas leaks from fuel elements. 14 figs.

  15. Radiation safety assessment of a system of small reactors for distributed energy.

    PubMed

    Odano, N; Ishida, T

    2005-01-01

    A passively safe small reactor for a distributed energy system, PSRD, is an integral type of light-water reactor with a thermal output of 100 or 300 MW aimed to be used for supplying district heat, electricity to small grids, and so on. Candidate locations for the PSRD as a distributed energy source are on-ground, deep underground, and in a seaside pit in the vicinity of the energy consumption area. Assessments of the radiation safety of a PSRD were carried out for three cases corresponding to normal operation, shutdown and a hypothetical postulated accident for several siting candidates. Results of the radiation safety assessment indicate that the PSRD design has sufficient shielding performance and capability and that the exposure to the general public is very low in the case of a hypothetical accident. PMID:16381690

  16. Expert system for identification of simultaneous and sequential reactor fuel failures with gas tagging

    DOEpatents

    Gross, Kenny C. (Bolingbrook, IL)

    1994-01-01

    Failure of a fuel element in a nuclear reactor core is determined by a gas tagging failure detection system and method. Failures are catalogued and characterized after the event so that samples of the reactor's cover gas are taken at regular intervals and analyzed by mass spectroscopy. Employing a first set of systematic heuristic rules which are applied in a transformed node space allows the number of node combinations which must be processed within a barycentric algorithm to be substantially reduced. A second set of heuristic rules treats the tag nodes of the most recent one or two leakers as "background" gases, further reducing the number of trial node combinations. Lastly, a "fuzzy" set theory formalism minimizes experimental uncertainties in the identification of the most likely volumes of tag gases. This approach allows for the identification of virtually any number of sequential leaks and up to five simultaneous gas leaks from fuel elements.

  17. Nuclear reactor fuel assembly duct-tube-to-inlet-nozzle attachment system

    DOEpatents

    Christiansen, David W. (Kennewick, WA); Smith, Bob G. (Kennewick, WA)

    1982-01-01

    A reusable system for removably attaching the lower end 21 of a nuclear reactor fuel assembly duct tube to an upper end 11 of a nuclear reactor fuel assembly inlet nozzle. The duct tube's lower end 21 has sides terminating in locking tabs 22 which end in inwardly-extending flanges 23. The flanges 23 engage recesses 13 in the top section 12 of the inlet nozzle's upper end 11. A retaining collar 30 slides over the inlet nozzle's upper end 11 to restrain the flanges 23 in the recesses 13. A locking nut 40 has an inside threaded portion 41 which engages an outside threaded portion 15 of the inlet nozzle's upper end 11 to secure the retaining collar 30 against protrusions 24 on the duct tube's sides.

  18. VICTORIA: A mechanistic model of radionuclide behavior in the reactor coolant system under severe accident conditions

    SciTech Connect

    Heames, T.J. (Science Applications International Corp., Albuquerque, NM (USA)); Williams, D.A.; Johns, N.A.; Chown, N.M. (UKAEA Atomic Energy Establishment, Winfrith (UK)); Bixler, N.E.; Grimley, A.J. (Sandia National Labs., Albuquerque, NM (USA)); Wheatley, C.J. (UKAEA Safety and Reliability Directorate, Culcheth (UK))

    1990-10-01

    This document provides a description of a model of the radionuclide behavior in the reactor coolant system (RCS) of a light water reactor during a severe accident. This document serves as the user's manual for the computer code called VICTORIA, based upon the model. The VICTORIA code predicts fission product release from the fuel, chemical reactions between fission products and structural materials, vapor and aerosol behavior, and fission product decay heating. This document provides a detailed description of each part of the implementation of the model into VICTORIA, the numerical algorithms used, and the correlations and thermochemical data necessary for determining a solution. A description of the code structure, input and output, and a sample problem are provided. The VICTORIA code was developed upon a CRAY-XMP at Sandia National Laboratories in the USA and a CRAY-2 and various SUN workstations at the Winfrith Technology Centre in England. 60 refs.

  19. Passive containment cooling system with drywell pressure regulation for boiling water reactor

    DOEpatents

    Hill, P.R.

    1994-12-27

    A boiling water reactor is described having a regulating valve for placing the wetwell in flow communication with an intake duct of the passive containment cooling system. This subsystem can be adjusted to maintain the drywell pressure at (or slightly below or above) wetwell pressure after the initial reactor blowdown transient is over. This addition to the PCCS design has the benefit of eliminating or minimizing steam leakage from the drywell to the wetwell in the longer-term post-LOCA time period and also minimizes the temperature difference between drywell and wetwell. This in turn reduces the rate of long-term pressure buildup of the containment, thereby extending the time to reach the design pressure limit. 4 figures.

  20. Passive containment cooling system with drywell pressure regulation for boiling water reactor

    DOEpatents

    Hill, Paul R. (Tucson, AZ)

    1994-01-01

    A boiling water reactor having a regulating valve for placing the wetwell in flow communication with an intake duct of the passive containment cooling system. This subsystem can be adjusted to maintain the drywell pressure at (or slightly below or above) wetwell pressure after the initial reactor blowdown transient is over. This addition to the PCCS design has the benefit of eliminating or minimizing steam leakage from the drywell to the wetwell in the longer-term post-LOCA time period and also minimizes the temperature difference between drywell and wetwell. This in turn reduces the rate of long-term pressure buildup of the containment, thereby extending the time to reach the design pressure limit.