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Sample records for heavy water reactor

  1. HEAVY WATER MODERATED NEUTRONIC REACTOR

    DOEpatents

    Szilard, L.

    1958-04-29

    A nuclear reactor of the type which utilizes uranium fuel elements and a liquid coolant is described. The fuel elements are in the form of elongated tubes and are disposed within outer tubes extending through a tank containing heavy water, which acts as a moderator. The ends of the fuel tubes are connected by inlet and discharge headers, and liquid bismuth is circulated between the headers and through the fuel tubes for cooling. Helium is circulated through the annular space between the outer tubes in the tank and the fuel tubes to cool the water moderator to prevent boiling. The fuel tubes are covered with a steel lining, and suitable control means, heat exchange means, and pumping means for the coolants are provided to complete the reactor assembly.

  2. HEAVY WATER COMPONENTS TEST REACTOR DECOMMISSIONING

    SciTech Connect

    Austin, W.; Brinkley, D.

    2011-10-13

    The Heavy Water Components Test Reactor (HWCTR) Decommissioning Project was initiated in 2009 as a Comprehensive Environmental Response, Compensation and Liability Act (CERCLA) Removal Action with funding from the American Recovery and Reinvestment Act (ARRA). This paper summarizes the history prior to 2009, the major D&D activities, and final end state of the facility at completion of decommissioning in June 2011. The HWCTR facility was built in 1961, operated from 1962 to 1964, and is located in the northwest quadrant of the Savannah River Site (SRS) approximately three miles from the site boundary. The HWCTR was a pressurized heavy water test reactor used to develop candidate fuel designs for heavy water power reactors. In December of 1964, operations were terminated and the facility was placed in a standby condition as a result of the decision by the U.S. Atomic Energy Commission to redirect research and development work on heavy water power reactors to reactors cooled with organic materials. For about one year, site personnel maintained the facility in a standby status, and then retired the reactor in place. In the early 1990s, DOE began planning to decommission HWCTR. Yet, in the face of new budget constraints, DOE deferred dismantlement and placed HWCTR in an extended surveillance and maintenance mode. The doors of the reactor facility were welded shut to protect workers and discourage intruders. In 2009 the $1.6 billion allocation from the ARRA to SRS for site footprint reduction at SRS reopened the doors to HWCTR - this time for final decommissioning. Alternative studies concluded that the most environmentally safe, cost effective option for final decommissioning was to remove the reactor vessel, both steam generators, and all equipment above grade including the dome. The transfer coffin, originally above grade, was to be placed in the cavity vacated by the reactor vessel and the remaining below grade spaces would be grouted. Once all above equipment

  3. METHOD OF OPERATING A HEAVY WATER MODERATED REACTOR

    DOEpatents

    Vernon, H.C.

    1962-08-14

    A method of removing fission products from the heavy water used in a slurry type nuclear reactor is described. According to the process the slurry is steam distilled with carbon tetrachloride so that at least a part of the heavy water and carbon tetrachloride are vaporized; the heavy water and carbon tetrachloride are separated; the carbon tetrachloride is returned to the steam distillation column at different points in the column to aid in depositing the slurry particles at the bottom of the column; and the heavy water portion of the condensate is purified. (AEC)

  4. Design of Recycle Pressurized Water Reactor with Heavy Water Moderation

    SciTech Connect

    Hibi, Koki; Uchita, Masato

    2004-03-15

    This study presents the conceptual design of the recycle pressurized water reactor (RPWR), which is an innovative PWR fueled with mixed oxide, moderated by heavy water, and having breeding ratios around 1.1. Most of the systems of RPWR can employ those of PWRs. The RPWR has no boric acid systems and has a small tritium removal system. The construction and operation costs would be similar to those of current PWRs. Heavy water cost has decreased drastically with up-to-date producing methods. The reliability of the systems of the RPWR is high, and the research and development cost for RPWR is very low because the core design is fundamentally based on the current PWR technology.

  5. Development of a Heavy Water Detritiation Plant for PIK Reactor

    SciTech Connect

    Alekseev, I.A.; Bondarenko, S.D.; Fedorchenko, O.A.; Konoplev, K.A.; Vasyanina, T.V.; Arkhipov, E.A.; Uborsky, V.V

    2005-07-15

    The research reactor PIK should be supplied with a Detritiation Plant (DP) to remove tritium from heavy water in order to reduce operator radiation dose and tritium emissions. The original design of the reactor PIK Detritiation Plant was completed several years ago. A number of investigations have been made to obtain data for the DP design. Nowadays the design of the DP is being revised on a basis of our investigations. The Combined Electrolysis and Catalytic Exchange (CECE) process will be used at the Detritiation Plant instead of Vapor Phase Catalytic Exchange. The experimental industrial plant for hydrogen isotope separation on the basis of the CECE process is under operation in Petersburg Nuclear Physics Institute. The plant was updated to provide a means for heavy water detritiation. Very high detritiation factors have been achieved in the plant. The use of the CECE process will allow the development of a more compact and less expensive detritiation plant for heavy water reactor PIK.

  6. Heavy Water Components Test Reactor Decommissioning - Major Component Removal

    SciTech Connect

    Austin, W.; Brinkley, D.

    2010-05-05

    The Heavy Water Components Test Reactor (HWCTR) facility (Figure 1) was built in 1961, operated from 1962 to 1964, and is located in the northwest quadrant of the Savannah River Site (SRS) approximately three miles from the site boundary. The HWCTR facility is on high, well-drained ground, about 30 meters above the water table. The HWCTR was a pressurized heavy water test reactor used to develop candidate fuel designs for heavy water power reactors. It was not a defense-related facility like the materials production reactors at SRS. The reactor was moderated with heavy water and was rated at 50 megawatts thermal power. In December of 1964, operations were terminated and the facility was placed in a standby condition as a result of the decision by the U.S. Atomic Energy Commission to redirect research and development work on heavy water power reactors to reactors cooled with organic materials. For about one year, site personnel maintained the facility in a standby status, and then retired the reactor in place. In 1965, fuel assemblies were removed, systems that contained heavy water were drained, fluid piping systems were drained, deenergized and disconnected and the spent fuel basin was drained and dried. The doors of the reactor facility were shut and it wasn't until 10 years later that decommissioning plans were considered and ultimately postponed due to budget constraints. In the early 1990s, DOE began planning to decommission HWCTR again. Yet, in the face of new budget constraints, DOE deferred dismantlement and placed HWCTR in an extended surveillance and maintenance mode. The doors of the reactor facility were welded shut to protect workers and discourage intruders. The $1.6 billion allocation from the American Recovery and Reinvestment Act to SRS for site clean up at SRS has opened the doors to the HWCTR again - this time for final decommissioning. During the lifetime of HWCTR, 36 different fuel assemblies were tested in the facility. Ten of these

  7. Recovery and packaging of tritium from Canadian heavy water reactors

    SciTech Connect

    Holtslander, W.J.; Goyette, V.; Harrison, T.E.; Miller, J.M.

    1985-09-01

    The Tritium Extraction Plant being built at Chalk River Nuclear Laboratories (CRNL) will be the first industrial scale demonstration of the Liquid Phase Catalytic Exchange (LPCE) process for transfer of tritium from heavy water to deuterium. The plant will also demonstrate new technology in the areas of electrolytic cells for D/sub 2/ generation, water cooled recombiners, metal hydride packaging and magnetically coupled blowers for tritium service. It will be used to detritiate the heavy water in Atomic Energy of Canada Limited's (AECL) reactors.

  8. Deployment Scenario of Heavy Water Cooled Thorium Breeder Reactor

    SciTech Connect

    Mardiansah, Deby; Takaki, Naoyuki

    2010-06-22

    Deployment scenario of heavy water cooled thorium breeder reactor has been studied. We have assumed to use plutonium and thorium oxide fuel in water cooled reactor to produce {sup 233}U which will be used in thorium breeder reactor. The objective is to analysis the potential of water cooled Th-Pu reactor for replacing all of current LWRs especially in Japan. In this paper, the standard Pressurize Water Reactor (PWR) has been designed to produce 3423 MWt; (i) Th-Pu PWR, (ii) Th-Pu HWR (MFR = 1.0) and (iii) Th-Pu HWR (MFR 1.2). The properties and performance of the core were investigated by using cell and core calculation code. Th-Pu PWR or HWR produces {sup 233}U to introduce thorium breeder reactor. The result showed that to replace all (60 GWe) LWR by thorium breeder reactor within a period of one century, Th-Pu oxide fueled PWR has insufficient capability to produce necessary amount of {sup 233}U and Th-Pu oxide fueled HWR has almost enough potential to produce {sup 233}U but shows positive void reactivity coefficient.

  9. CHARACTERIZATION OF RADIOACTIVITY IN THE REACTOR VESSEL OF THE HEAVY WATER COMPONENT TEST REACTOR

    SciTech Connect

    Vinson, Dennis

    2010-06-01

    The Heavy Water Component Test Reactor (HWCTR) facility is a pressurized heavy water reactor that was used to test candidate fuel designs for heavy water power reactors. The reactor operated at nominal power of 50 MW{sub th}. The reactor coolant loop operated at 1200 psig and 250 C. Two isolated test loop were designed into the reactor to provide special test conditions. Fig. 1 shows a cut-away view of the reactor. The two loops are contained in four inch diameter stainless steel piping. The HWCTR was operated for only a short duration, from March 1962 to December 1964 in order to test the viability of test fuel elements and other reactor components for use in a heavy water power reactor. The reactor achieved 13,882 MWd of total power while testing 36 different fuel assemblies. In the course of operation, HWCTR experienced the cladding failures of 10 separate test fuel assemblies. In each case, the cladding was breached with some release of fuel core material into the isolated test loop, causing fission product and actinide contamination in the main coolant loop and the liquid and boiling test loops. Despite the contribution of the contamination from the failed fuel, the primary source of radioactivity in the HWCTR vessel and internals is the activation products in the thermal shields, and to a lesser degree, activation products in the reactor vessel walls and liner. A detailed facility characterization report of the HWCTR facility was completed in 1996. Many of the inputs and assumptions in the 1996 characterization report were derived from the HWCTR decommissioning plan published in 1975. The current paper provides an updated assessment of the radioisotopic characteristics of the HWCTR vessel and internals to support decommissioning activities on the facility.

  10. Accident analysis of heavy water cooled thorium breeder reactor

    SciTech Connect

    Yulianti, Yanti; Su’ud, Zaki; Takaki, Naoyuki

    2015-04-16

    Thorium has lately attracted considerable attention because it is accumulating as a by-product of large scale rare earth mining. The objective of research is to analyze transient behavior of a heavy water cooled thorium breeder that is designed by Tokai University and Tokyo Institute of Technology. That is oxide fueled, PWR type reactor with heavy water as primary coolant. An example of the optimized core has relatively small moderator to fuel volume ratio (MFR) of 0.6 and the characteristics of the core are burn-up of 67 GWd/t, breeding ratio of 1.08, burn-up reactivity loss during cycles of < 0.2% dk/k, and negative coolant reactivity coefficient. One of the nuclear reactor accidents types examined here is Unprotected Transient over Power (UTOP) due to withdrawing of the control rod that result in the positive reactivity insertion so that the reactor power will increase rapidly. Another accident type is Unprotected Loss of Flow (ULOF) that caused by failure of coolant pumps. To analyze the reactor accidents, neutron distribution calculation in the nuclear reactor is the most important factor. The best expression for the neutron distribution is the Boltzmann transport equation. However, solving this equation is very difficult so that the space-time diffusion equation is commonly used. Usually, space-time diffusion equation is solved by employing a point kinetics approach. However, this approach is less accurate for a spatially heterogeneous nuclear reactor and the nuclear reactor with quite large reactivity input. Direct method is therefore used to solve space-time diffusion equation which consider spatial factor in detail during nuclear reactor accident simulation. Set of equations that obtained from full implicit finite-difference method is solved by using iterative methods. The indication of UTOP accident is decreasing macroscopic absorption cross-section that results large external reactivity, and ULOF accident is indicated by decreasing coolant flow. The

  11. Accident analysis of heavy water cooled thorium breeder reactor

    NASA Astrophysics Data System (ADS)

    Yulianti, Yanti; Su'ud, Zaki; Takaki, Naoyuki

    2015-04-01

    Thorium has lately attracted considerable attention because it is accumulating as a by-product of large scale rare earth mining. The objective of research is to analyze transient behavior of a heavy water cooled thorium breeder that is designed by Tokai University and Tokyo Institute of Technology. That is oxide fueled, PWR type reactor with heavy water as primary coolant. An example of the optimized core has relatively small moderator to fuel volume ratio (MFR) of 0.6 and the characteristics of the core are burn-up of 67 GWd/t, breeding ratio of 1.08, burn-up reactivity loss during cycles of < 0.2% dk/k, and negative coolant reactivity coefficient. One of the nuclear reactor accidents types examined here is Unprotected Transient over Power (UTOP) due to withdrawing of the control rod that result in the positive reactivity insertion so that the reactor power will increase rapidly. Another accident type is Unprotected Loss of Flow (ULOF) that caused by failure of coolant pumps. To analyze the reactor accidents, neutron distribution calculation in the nuclear reactor is the most important factor. The best expression for the neutron distribution is the Boltzmann transport equation. However, solving this equation is very difficult so that the space-time diffusion equation is commonly used. Usually, space-time diffusion equation is solved by employing a point kinetics approach. However, this approach is less accurate for a spatially heterogeneous nuclear reactor and the nuclear reactor with quite large reactivity input. Direct method is therefore used to solve space-time diffusion equation which consider spatial factor in detail during nuclear reactor accident simulation. Set of equations that obtained from full implicit finite-difference method is solved by using iterative methods. The indication of UTOP accident is decreasing macroscopic absorption cross-section that results large external reactivity, and ULOF accident is indicated by decreasing coolant flow. The

  12. Toxicity of irradiated advanced heavy water reactor fuels.

    PubMed

    Priest, N D; Richardson, R B; Edwards, G W R

    2013-02-01

    The good neutron economy and online refueling capability of the CANDU® heavy water moderated reactor (HWR) enable it to use many different fuels such as low enriched uranium (LEU), plutonium, or thorium, in addition to its traditional natural uranium (NU) fuel. The toxicity and radiological protection methods for these proposed fuels, unlike those for NU, are not well established. This study uses software to compare the fuel composition and toxicity of irradiated NU fuel against those of two irradiated advanced HWR fuel bundles as a function of post-irradiation time. The first bundle investigated is a CANFLEX® low void reactor fuel (LVRF), of which only the dysprosium-poisoned central element, and not the outer 42 LEU elements, is specifically analyzed. The second bundle investigated is a heterogeneous high-burnup (LEU,Th)O(2) fuelled bundle, whose two components (LEU in the outer 35 elements and thorium in the central eight elements) are analyzed separately. The LVRF central element was estimated to have a much lower toxicity than that of NU at all times after shutdown. Both the high burnup LEU and the thorium fuel had similar toxicity to NU at shutdown, but due to the creation of such inhalation hazards as (238)Pu, (240)Pu, (242)Am, (242)Cm, and (244)Cm (in high burnup LEU), and (232)U and (228)Th (in irradiated thorium), the toxicity of these fuels was almost double that of irradiated NU after 2,700 d of cooling. New urine bioassay methods for higher actinoids and the analysis of thorium in fecal samples are recommended to assess the internal dose from these two fuels.

  13. Accident management for indian pressurized heavy water reactors

    SciTech Connect

    Hajela, S.; Grover, R.; Ghadge, S.G.; Bajaj, S.S.

    2006-07-01

    Indian nuclear power program as of now is mainly based on Pressurized Heavy Water Reactors (PHWRs). Operating Procedures for normal power operation and Emergency Operating Procedures for operational transients and accidents within design basis exist for all Indian PHWRs. In addition, on-site and off-site emergency response procedures are also available for these NPPs. The guidelines needed for severe accidents mitigation are now formally being documented for Indian PHWRs. Also, in line with International trend of having symptom based emergency handling, the work is in advanced stage for preparation of symptom-based emergency operating procedures. Following a plant upset condition; a number of alarms distributed in different information systems appear in the control room to aid operator to identify the nature of the event. After identifying the event, appropriate intervention in the form of event based emergency operating procedure is put into use by the operating staff. However, if the initiating event cannot be unambiguously identified or after the initial event some other failures take place, then the selected event based emergency operating procedure will not be optimal. In such a case, reactor safety is ensured by monitoring safety functions (depicted by selected plant parameters grouped together) throughout the event handling so that the barriers to radioactivity release namely, fuel and fuel cladding, primary heat transport system integrity and containment remain intact. Simultaneous monitoring of all these safety functions is proposed through status trees and this concept will be implemented through a computer-based system. For beyond design basis accidents, event sequences are identified which may lead to severe core damage. As part of this project, severe accident mitigation guidelines are being finalized for the selected event sequences. The paper brings out the details of work being carried out for Indian PHWRs for symptom based event handling and severe

  14. Analysis of Removal Alternatives for the Heavy Water Components Test Reactor at the Savannah River Site

    SciTech Connect

    Owen, M.B.

    1996-08-01

    This engineering study was developed to evaluate different options for decommissioning of the Heavy Water Components Test Reactor (HWCTR) at the Savannah River Site. This document will be placed in the DOE-SRS Area reading rooms for a period of 30 days in order to obtain public input to plans for the demolition of HWCTR.

  15. Study on hydrodynamically induced dryout and post dryout important to heavy water reactors

    SciTech Connect

    Ishii, M.; Revankar, S.T.; Nair, S.; Lele, S.; Eberle, C.S.; Babelli, I.

    1995-01-01

    Recently, the safety of low pressure liquid cooled nuclear reactors has become a very important issue with reference to the operation of the heavy water reactors at Savannah River Plant. Under accident conditions such as loss-of-flow or loss-of-coolant, these reactors typically encounter unstable two-phase flow which may lead to the occurrence of dryout and subsequent fuel failure. An analytical study using the one-dimensional drift flux model was carried out to investigate the two-phase flow instability for Westinghouse Savannah River Site reactor. The analysis indicates that the first and higher order instabilities exist in the possible transient operational conditions. The instabilities are encountered at higher heat fluxes or lower flow rates. The subcooling has a stabilizing effect except at very low subcooling. An experimental loop has been designed and constructed. A study was conducted on the CHF induced by various flow instabilities. Details of this test loop are presented.

  16. Pressurized heavy water reactor fuel behaviour in power ramp conditions

    NASA Astrophysics Data System (ADS)

    Ionescu, S.; Uţă, O.; Pârvan, M.; Ohâi, D.

    2009-03-01

    In order to check and improve the quality of the Romanian CANDU fuel, an assembly of six CANDU fuel rods has been subjected to a power ramping test in the 14 MW TRIGA reactor at INR. After testing, the fuel rods have been examined in the hot cells using post-irradiation examination (PIE) techniques such as: visual inspection and photography, eddy current testing, profilometry, gamma scanning, fission gas release and analysis, metallography, ceramography, burn-up determination by mass spectrometry, mechanical testing. This paper describes the PIE results from one out of the six fuel rods. The PIE results concerning the integrity, dimensional changes, oxidation, hydriding and mechanical properties of the sheath, the fission-products activity distribution in the fuel column, the pressure, volume and composition of the fission gas, the burn-up, the isotopic composition and structural changes of the fuel enabled the characterization of the behaviour of the Romanian CANDU fuel in power ramping conditions performed in the TRIGA materials testing reactor.

  17. Fission product release phenomena during core melt accidents in metal fueled heavy water reactors

    SciTech Connect

    Ellison, P G; Hyder, M L; Monson, P R; Randolph, H W; Hagrman, D L; McClure, P R; Leonard, M T

    1990-01-01

    The phenomena that determine fission product release rates from a core melting accident in a metal-fueled, heavy water reactor are described in this paper. This information is obtained from the analysis of the current metal fuel experimental data base and from the results of analytical calculations. Experimental programs in place at the Savannah River Site are described that will provide information to resolve uncertainties in the data base. The results of the experiments will be incorporated into new severe accident computer codes recently developed for this reactor design. 47 refs., 4 figs.

  18. The Possible Use of Cermet Fuel in the Dido and Pluto Heavy-Water Research Reactors,

    DTIC Science & Technology

    1981-08-01

    tions, I.A.E.A., Vienna (September 1980). 5. TOULOUKIAN , Y.S., ’Thermophysical properties of high-temperature solid materials’, Purdue University...the implications of using lower-enrichment fuels in heavy-water-moderated research reactors. The uranium content of the U/Al alloy at present used...cannot be increased sufficiently to maintain the requisite UL3 5 content without undesirable effects on the physical properties of the alloy . A different

  19. Radio-toxicity of spent fuel of the advanced heavy water reactor.

    PubMed

    Anand, S; Singh, K D S; Sharma, V K

    2010-01-01

    The Advanced Heavy Water Reactor (AHWR) is a new power reactor concept being developed at Bhabha Atomic Research Centre, Mumbai. The reactor retains many desirable features of the existing Pressurised Heavy Water Reactor (PHWR), while incorporating new, advanced safety features. The reactor aims to utilise the vast thorium resources available in India. The reactor core will use plutonium as the make-up fuel, while breeding (233)U in situ. On account of this unique combination of fuel materials, the operational characteristics of the fuel as determined by its radioactivity, decay heat and radio-toxicity are being viewed with great interest. Radio-toxicity of the spent fuel is a measure of potential radiological hazard to the members of the public and also important from the ecological point of view. The radio-toxicity of the AHWR fuel is extremely high to start with, being approximately 10(4) times that of the fresh natural U fuel used in a PHWR, and continues to remain relatively high during operation and subsequent cooling. A unique feature of this fuel is the peak observed in its radio-toxicity at approximately 10(5) y of decay cooling. The delayed increase in fuel toxicity has been traced primarily to a build-up of (229)Th, (230)Th and (226)Ra. This phenomenon has been observed earlier for thorium-based fuels and is confirmed for the AHWR fuel. This paper presents radio-toxicity data for AHWR spent fuel up to a period of 10(6) y and the results are compared with the radio-toxicity of PHWR.

  20. CECE alternative for upgrading/detritiation in heavy water nuclear reactors and for tritium recovery in fusion reactors

    SciTech Connect

    Spagnolo, D.A.; Miller, A.I.

    1995-10-01

    The Combined Electrolysis Catalytic Exchange (CECE) process, utilizing AECL`s wetproofed catalyst, is ideally suited for extracting tritium from water because of its high isotopic separation factor and near-ambient operating conditions. Several CECE options are compared with the more conventional DW-VPCE arrangements for heavy water upgrading and detritiation of CANDU nuclear reactors and for detritiation of fusion facilities such as ITER. For both applications, CECE offers a more economical alternative over conventional technology. Experimental data on catalyst activity and lifetime are also presented and past commercial applications of the AECL catalyst are reviewed. AECL has recently committed to assembly of a CECE upgrading/detritiation demonstration facility. 15 refs., 5 figs., 1 tab.

  1. Heavy metal removal by combining anaerobic upflow packed bed reactors with water hyacinth ponds.

    PubMed

    Sekomo, Christian Birame; Kagisha, Vedaste; Rousseau, Diederik; Lens, Piet

    2012-06-01

    The removal of four selected heavy metals (Cu, Cd, Pb and Zn) has been assessed in an upflow anaerobic packed bed reactor filled with porous volcanic rock as an adsorbent and an attachment surface for bacterial growth. Two different feeding regimes were applied using low (5 mg L(-1) of heavy metal each) and high (10 mg L(-1) of heavy metal each) strength wastewater. After a start-up and acclimatization period of 44 days, each regime was operated for a period of 10 days with a hydraulic retention time of one day. Good removal efficiencies of at least 86% were achieved for both the low and high strength wastewater. A subsequent water hyacinth pond with a hydraulic retention time of one day removed an additional 61% Cd, 59% Cu, 49% Pb and 42% Zn, showing its importance as a polishing step. The water hyacinth plant in the post-treatment step accumulated heavy metals mainly in the root system. Overall metal removal efficiencies at the outlet of the integrated system were 98% for Cd, 99% for Cu, 98% for Pb and 84% for Zn. Therefore, the integrated system can be used as an alternative treatment system for metal-polluted wastewater, especially in developing countries.

  2. Fracture mechanics and full scale pipe break testing for the Department of Environment's New Production Reactor-Heavy Water Reactor

    NASA Astrophysics Data System (ADS)

    Poole, A. B.

    Oak Ridge National Laboratory (ORNL) is completing a major task for the Department of Energy (DOE) in the demonstration that the primary piping of the New Production Heavy Water Reactor (NPR-HWR), with its relatively moderate temperature and pressure, should not suffer an instantaneous Double-Ended-Guillotine-Break (DEGB) under design basis loadings and conditions. The growth of possible small preexisting defects in the piping wall was estimated over a plant life of 60 years. This worst case flaw was then evaluated using fracture mechanics methods. J estimation methods and tearing instability approximations used in this analysis are discussed in this paper. It was established that this worst case flaw would increase in size by at least 14 times before pipe instability during an earthquake would even begin to be possible. The fatigue crack growth analysis is discussed in this paper.

  3. FMDP Reactor Alternative Summary Report: Volume 2 - CANDU heavy water reactor alternative

    SciTech Connect

    Greene, S.R.; Spellman, D.J.; Bevard, B.B.

    1996-09-01

    The Department of Energy Office of Fissile Materials Disposition (DOE/MD) initiated a detailed analysis activity to evaluate each of ten plutonium disposition alternatives that survived an initial screening process. This document, Volume 2 of a four volume report, summarizes the results of these analyses for the CANDU reactor based plutonium disposition alternative.

  4. A Diagnostic Hierarchy Approach to Root Cause Analysis for Heavy Water Reactor Malfunction Management

    SciTech Connect

    Miller, D.W.; Hajek, B.K.; Hines, J.W.

    1993-10-30

    The Nuclear Engineering and Chemical Engineering Artificial Intelligence Groups at The Ohio State University have developed a diagnostic system for the heavy water production reactors at the Savannah River Site. The diagnostic module of the system uses hybrid hierarchical decomposition methodology to decompose the search space. The knowledge is arranged so that the search space is traversed similarly to how an expert would solve the problem. The system was tested on the SRS development simulator and the results show that the system can properly diagnose all the process water and cooling water malfunctions that are programmed into the simulator. The system was not validated by operators due to hardware unavailability. Since the New Production Reactor development efforts have been halted, the probability for future work on this project is unlikely. The development used a standardized Verification and Validation program to assist in the design and construction of the system. The use of this standardized procedure is referred to as a text book example of designing an expert system in the expectation that its use would provide guidance in future projects. Of the eight phases of the software development lifecycle, five of the phases were completed and documented.

  5. Modifications to MELCOR for the analysis of heavy-water moderated, U-A1 fuel reactors

    SciTech Connect

    Church, J.P. ); Leonard, M.T.; Williams, K.A. )

    1990-01-01

    The MELCOR computer code is being used as the point of departure to develop an integrated severe accident analysis computer code for the heavy-water moderated U-Al fuel reactors. The resulting computer code (MELCOR/SR) provides a practical and comprehensive analytical tool for evaluating severe accident behavior in the Savannah River Site (SRS) production reactors. The technical scope of this development effort is summarized in this paper. Other companion papers are cited that provide additional details regarding particular models.

  6. Minor actinide transmutation in thorium and uranium matrices in heavy water moderated reactors

    SciTech Connect

    Bhatti, Zaki; Hyland, B.; Edwards, G.W.R.

    2013-07-01

    The irradiation of Th{sup 232} breeds fewer of the problematic minor actinides (Np, Am, Cm) than the irradiation of U{sup 238}. This characteristic makes thorium an attractive potential matrix for the transmutation of these minor actinides, as these species can be transmuted without the creation of new actinides as is the case with a uranium fuel matrix. Minor actinides are the main contributors to long term decay heat and radiotoxicity of spent fuel, so reducing their concentration can greatly increase the capacity of a long term deep geological repository. Mixing minor actinides with thorium, three times more common in the Earth's crust than natural uranium, has the additional advantage of improving the sustainability of the fuel cycle. In this work, lattice cell calculations have been performed to determine the results of transmuting minor actinides from light water reactor spent fuel in a thorium matrix. 15-year-cooled group-extracted transuranic elements (Np, Pu, Am, Cm) from light water reactor (LWR) spent fuel were used as the fissile component in a thorium-based fuel in a heavy water moderated reactor (HWR). The minor actinide (MA) transmutation rates, spent fuel activity, decay heat and radiotoxicity, are compared with those obtained when the MA were mixed instead with natural uranium and taken to the same burnup. Each bundle contained a central pin containing a burnable neutron absorber whose initial concentration was adjusted to have the same reactivity response (in units of the delayed neutron fraction β) for coolant voiding as standard NU fuel. (authors)

  7. Residual radioactivity guidelines for the heavy water components test reactor at the Savannah River Site

    SciTech Connect

    Owen, M.B. Smith, R.; McNeil, J.

    1997-04-01

    Guidelines were developed for acceptable levels of residual radioactivity in the Heavy Water Components Test Reactor (HWCTR) facility at the conclusion of its decommissioning. Using source terms developed from data generated in a detailed characterization study, the RESRAD and RASRAD-BUILD computer codes were used to calculate derived concentration guideline levels (DCGLs) for the radionuclides that will remain in the facility. The calculated DCGLs, when compared to existing concentrations of radionuclides measured during a 1996 characterization program, indicate that no decontamination of concrete surfaces will be necessary. Also, based on the results of the calculations, activated concrete in the reactor biological shield does not have to be removed, and imbedded radioactive piping in the facility can remain in place. Viewed in another way, the results of the calculations showed that the present inventory of residual radioactivity in the facility (not including that associated with the reactor vessel and steam generators) would produce less than one millirem per year above background to a hypothetical individual on the property. The residual radioactivity is estimated to be approximately 0.04 percent of the total inventory in the facility as of March, 1997. According to the results, the only radionuclides that would produce greater than 0.0.1-millirem per year are Am-241 (0.013 mrem/yr at 300 years), C-14 (0.022 mrem/yr at 1000 years) and U-238 (0.034 mrem/yr at 6000 years). Human exposure would occur only through the groundwater pathways, that is, from water drawn from, a well on the property. The maximum exposure would be approximately one percent of the 4 millirem per year ground water exposure limit established by the U.S. Environmental Protection Agency. 11 refs., 13 figs., 15 tabs.

  8. Hydrodynamically induced dryout and post dryout important to heavy water reactors: A yearly progress report

    SciTech Connect

    Ishii, M.; Revankar, S.T.; Babelli, I.; Lele, S.

    1992-06-01

    Recently, the safety of low pressure liquid cooled nuclear reactors has become a very important issue with reference to the operation of the heavy water reactors at Savannah River Plant. Under accident conditions such as loss-of-flow or loss-of-coolant, these reactors typically encounter unstable two-phase flow which may lead to the occurrence of dryout and subsequent fuel failure. An analytical study using the one-dimensional drift flux model was carried out to investigate the two-phase flow instability for Westinghouse Savannah River Site reactor. The analysis indicates that the first and higher order instabilities exist in the possible transient operational conditions. The instabilities are encountered at higher heat fluxes or lower flow rates. The subcooling has a stabilizing effect except at very low subcooling. An experimental loop has been designed and constructed to study the CBF induced by various flow instabilities. Details of this test loop are presented. Initial test results have been presented. The two-phase flow regimes and hydrodynamic behaviors in the post dryout region have been studied under propagating rewetting conditions. The effect of subcooling and inlet velocity on flow transition as well as on the quench front propagation was investigated. The test liquid was Freon 113 which was introduced into the bottom of the quartz test section whose walls were maintained well above the film boiling temperature of the test liquid, via a transparent heat transfer fluid. The flow regimes observed down stream of the upward moving quench front were the rough wavy, the agitated, and the dispersed droplet/ligaments. A correlation for the flow regime transition between the inverted annular and the dispersed droplet/ligament flow patterns was developed. The correlation showed a marked dependence on the void fraction at the CBF location and hence on the flow regime encountered in the pre-CBF region.

  9. Core loading pattern optimization of thorium fueled heavy water breeder reactor using genetic algorithm

    SciTech Connect

    Soewono, C. N.; Takaki, N.

    2012-07-01

    In this work genetic algorithm was proposed to solve fuel loading pattern optimization problem in thorium fueled heavy water reactor. The objective function of optimization was to maximize the conversion ratio and minimize power peaking factor. Those objectives were simultaneously optimized using non-dominated Pareto-based population ranking optimal method. Members of non-dominated population were assigned selection probabilities based on their rankings in a manner similar to Baker's single criterion ranking selection procedure. A selected non-dominated member was bred through simple mutation or one-point crossover process to produce a new member. The genetic algorithm program was developed in FORTRAN 90 while neutronic calculation and analysis was done by COREBN code, a module of core burn-up calculation for SRAC. (authors)

  10. Inhalation radiotoxicity of irradiated thorium as a heavy water reactor fuel

    SciTech Connect

    Edwards, G.W.R.; Priest, N.D.; Richardson, R.B.

    2013-07-01

    The online refueling capability of Heavy Water Reactors (HWRs), and their good neutron economy, allows a relatively high amount of neutron absorption in breeding materials to occur during normal fuel irradiation. This characteristic makes HWRs uniquely suited to the extraction of energy from thorium. In Canada, the toxicity and radiological protection methods dealing with personnel exposure to natural uranium (NU) spent fuel (SF) are well-established, but the corresponding methods for irradiated thorium fuel are not well known. This study uses software to compare the activity and toxicity of irradiated thorium fuel ('thorium SF') against those of NU. Thorium elements, contained in the inner eight elements of a heterogeneous high-burnup bundle having LEU (Low-enriched uranium) in the outer 35 elements, achieve a similar burnup to NU SF during its residence in a reactor, and the radiotoxicity due to fission products was found to be similar. However, due to the creation of such inhalation hazards as U-232 and Th-228, the radiotoxicity of thorium SF was almost double that of NU SF after sufficient time has passed for the decay of shorter-lived fission products. Current radio-protection methods for NU SF exposure are likely inadequate to estimate the internal dose to personnel to thorium SF, and an analysis of thorium in fecal samples is recommended to assess the internal dose from exposure to this fuel. (authors)

  11. Annular seed-blanket thorium fuel core concepts for heavy water moderated reactors

    SciTech Connect

    Bromley, B.P.; Hyland, B.

    2013-07-01

    New reactor concepts to implement thorium-based fuel cycles have been explored to achieve maximum resource utilization. Pressure tube heavy water reactors (PT-HWR) are highly advantageous for implementing the use of thorium-based fuels because of their high neutron economy and on-line re-fuelling capability. The use of heterogeneous seed-blanket core concepts in a PT-HWR where higher-fissile-content seed fuel bundles are physically separate from lower-fissile-content blanket bundles allows more flexibility and control in fuel management to maximize the fissile utilization and conversion of fertile fuel. The lattice concept chosen is a 35-element bundle made with a homogeneous mixture of reactor grade Pu and Th, and with a central zirconia rod to help reduce coolant void reactivity. Several annular heterogeneous seed-blanket core concepts with plutonium-thorium-based fuels in a 700-MWe-class PT-HWR were analyzed, using a once-through thorium (OTT) cycle. Different combinations of seed and blanket fuel were tested to determine the impact on core-average burnup, fissile utilization, power distributions, and other performance parameters. It was found that the various core concepts can achieve a fissile utilization that is up to 30% higher than is currently achieved in a PT-HWR using conventional natural uranium fuel bundles. Up to 67% of the Pu is consumed; up to 43% of the energy is produced from thorium, and up to 363 kg/year of U-233 is produced. Seed-blanket cores with ∼50% content of low-power blanket bundles may require power de-rating (∼58% to 65%) to avoid exceeding maximum limits for peak channel power, bundle power and linear element ratings. (authors)

  12. Checkerboard seed-blanket thorium fuel core concepts for heavy water moderated reactors

    SciTech Connect

    Bromley, B.P.; Hyland, B.

    2013-07-01

    New reactor concepts to implement thorium-based fuel cycles have been explored to achieve maximum resource utilization. Pressure tube heavy water reactors (PT-HWR) are highly advantageous for implementing the use of thorium-based fuels because of their high neutron economy and on-line re-fuelling capability. The use of heterogeneous seed-blanket core concepts in a PT-HWR where higher-fissile-content seed fuel bundles are physically separate from lower-fissile-content blanket bundles allows more flexibility and control in fuel management to maximize the fissile utilization and conversion of fertile fuel. The lattice concept chosen was a 35-element bundle made with a homogeneous mixture of reactor grade Pu (about 67 wt% fissile) and Th, and with a central zirconia rod to help reduce coolant void reactivity. Several checkerboard heterogeneous seed-blanket core concepts with plutonium-thorium-based fuels in a 700-MWe-class PT-HWR were analyzed, using a once-through thorium (OTT) cycle. Different combinations of seed and blanket fuel were tested to determine the impact on core-average burnup, fissile utilization, power distributions, and other performance parameters. It was found that various checkerboard core concepts can achieve a fissile utilization that is up to 26% higher than that achieved in a PT-HWR using more conventional natural uranium fuel bundles. Up to 60% of the Pu is consumed; up to 43% of the energy is produced from thorium, and up to 303 kg/year of Pa-233/U-233/U-235 are produced. Checkerboard cores with about 50% of low-power blanket bundles may require power de-rating (65% to 74%) to avoid exceeding maximum limits for channel and bundle powers and linear element ratings. (authors)

  13. Lattice cell and full core physics of internally cooled annular fuel in heavy water moderated reactors

    SciTech Connect

    Armstrong, J.; Hamilton, H.; Hyland, B.

    2013-07-01

    A program is underway at Atomic Energy of Canada Limited (AECL) to develop a new fuel bundle concept to enable greater burnups for PT-HWR (pressure tube heavy water reactor) cores. One option that AECL is investigating is an internally cooled annular fuel (ICAF) element concept. ICAF contains annular cylindrical pellets with cladding on the inner and outer diameters. Coolant flows along the outside of the element and through the centre. With such a concept, the maximum fuel temperature as a function of linear element rating is significantly reduced compared to conventional, solid-rod type fuel. The preliminary ICAF bundle concept considered in this study contains 24 half-metre long internally cooled annular fuel elements and one non-fuelled centre pin. The introduction of the non-fuelled centre pin reduces the coolant void reactivity (CVR), which is the increase in reactivity that occurs on voiding the coolant in accident scenarios. Lattice cell and full core physics calculations of the preliminary ICAF fuel bundle concept have been performed for medium burnups of approximately 18 GWd/tU using WIMS-AECL and reactor fuel simulation program (RFSP). The results will be used to assist in concept configuration optimization. The effects of radial and axial core power distributions, linear element power ratings, refuelling rates and operational power ramps have been analyzed. The results suggest that burnups of greater than 18 GWd/tU can be achieved in current reactor designs. At approximately 18 GWd/tU, expected maximum linear element ratings in a PT-HWR with online-refuelling are approximately 90 kW/m. These conditions would be prohibitive for solid-rod fuel, but may be possible in ICAF fuel given the reduced maximum fuel temperature as a function of linear element rating. (authors)

  14. Further evaluations of the toxicity of irradiated advanced heavy water reactor fuels.

    PubMed

    Edwards, Geoffrey W R; Priest, Nicholas D

    2014-11-01

    The neutron economy and online refueling capability of heavy water moderated reactors enable them to use many different fuel types, such as low enriched uranium, plutonium mixed with uranium, or plutonium and/or U mixed with thorium, in addition to their traditional natural uranium fuel. However, the toxicity and radiological protection methods for fuels other than natural uranium are not well established. A previous paper by the current authors compared the composition and toxicity of irradiated natural uranium to that of three potential advanced heavy water fuels not containing plutonium, and this work uses the same method to compare irradiated natural uranium to three other fuels that do contain plutonium in their initial composition. All three of the new fuels are assumed to incorporate plutonium isotopes characteristic of those that would be recovered from light water reactor fuel via reprocessing. The first fuel investigated is a homogeneous thorium-plutonium fuel designed for a once-through fuel cycle without reprocessing. The second fuel is a heterogeneous thorium-plutonium-U bundle, with graded enrichments of U in different parts of a single fuel assembly. This fuel is assumed to be part of a recycling scenario in which U from previously irradiated fuel is recovered. The third fuel is one in which plutonium and Am are mixed with natural uranium. Each of these fuels, because of the presence of plutonium in the initial composition, is determined to be considerably more radiotoxic than is standard natural uranium. Canadian nuclear safety regulations require that techniques be available for the measurement of 1 mSv of committed effective dose after exposure to irradiated fuel. For natural uranium fuel, the isotope Pu is a significant contributor to the committed effective dose after exposure, and thermal ionization mass spectrometry is sensitive enough that the amount of Pu excreted in urine is sufficient to estimate internal doses, from all isotopes, as low

  15. Fracture analysis of axially cracked pressure tube of pressurized heavy water reactor

    SciTech Connect

    Krishnan, S.; Bhasin, V.; Mahajan, S.C.

    1997-04-01

    Three Dimensional (313) finite element elastic plastic fracture analysis was done for through wall axially cracked thin pressure tubes of 220 MWe Indian Pressurized Heavy Water Reactor. The analysis was done for Zr-2 and Zr-2.5Nb pressure tubes operating at 300{degrees}C and subjected to 9.5 Mpa internal pressure. Critical crack length was determined based on tearing instability concept. The analysis included the effect of crack face pressure due to the leaking fluid from tube. This effect was found to be significant for pressure tubes. The available formulae for calculating J (for axially cracked tubes) do not take into account the effect of crack face pressure. 3D finite element analysis also gives insight into variation of J across the thickness of pressure tube. It was observed that J is highest at the mid-surface of tube. The results have been presented in the form of across the thickness average J value and a peak factor on J. Peak factor on J is ratio of J at mid surface to average J value. Crack opening area for different cracked lengths was calculated from finite element results. The fracture assessment of pressure tubes was also done using Central Electricity Generating Board R-6 method. Ductile tearing was considered.

  16. Analysis of removal alternatives for the Heavy Water Components Test Reactor at the Savannah River Site. Revision 1

    SciTech Connect

    Owen, M.B.

    1997-04-01

    This engineering study evaluates different alternatives for decontamination and decommissioning of the Heavy Water Components Test Reactor (HWCTR). Cooled and moderated with pressurized heavy water, this uranium-fueled nuclear reactor was designed to test fuel assemblies for heavy water power reactors. It was operated for this purpose from march of 1962 until December of 1964. Four alternatives studied in detail include: (1) dismantlement, in which all radioactive and hazardous contaminants would be removed, the containment dome dismantled and the property restored to a condition similar to its original preconstruction state; (2) partial dismantlement and interim safe storage, where radioactive equipment except for the reactor vessel and steam generators would be removed, along with hazardous materials, and the building sealed with remote monitoring equipment in place to permit limited inspections at five-year intervals; (3) conversion for beneficial reuse, in which most radioactive equipment and hazardous materials would be removed and the containment building converted to another use such as a storage facility for radioactive materials, and (4) entombment, which involves removing hazardous materials, filling the below-ground structure with concrete, removing the containment dome and pouring a concrete cap on the tomb. Also considered was safe storage, but this approach, which has, in effect, been followed for the past 30 years, did not warrant detailed evaluation. The four other alternatives were evaluate, taking into account factors such as potential effects on the environment, risks, effectiveness, ease of implementation and cost. The preferred alternative was determined to be dismantlement. This approach is recommended because it ranks highest in the comparative analysis, would serve as the best prototype for the site reactor decommissioning program and would be most compatible with site property reuse plans for the future.

  17. Evaluate the radioactivity along the central thimble hole of a decommissioned heavy water research reactor using TLD approach.

    PubMed

    Lee, Lun-Hui; Sher, Hai-Feng; Lu, I-Hsin; Pan, Lung-Kwang

    2012-04-01

    The radioactivity along the central thimble hole of a decommissioned heavy water research reactor, TRR, was evaluated using TLD approach. The decay radionuclide was verified to be Co-60. The dose along the TRR central thimble hole was detected and revised by performing an unfolding analysis. The revised data reduced to 70-90% of the original data (for example, the maximum dose rate was reduced from 6447 to 4831 mSv/h,) and were more reliable. Copyright © 2011 Elsevier Ltd. All rights reserved.

  18. An economic analysis of a light and heavy water moderated reactor synergy: burning americium using recycled uranium

    SciTech Connect

    Wojtaszek, D.; Edwards, G.

    2013-07-01

    An economic analysis is presented for a proposed synergistic system between 2 nuclear utilities, one operating light water reactors (LWR) and another running a fleet of heavy water moderated reactors (HWR). Americium is partitioned from LWR spent nuclear fuel (SNF) to be transmuted in HWRs, with a consequent averted disposal cost to the LWR operator. In return, reprocessed uranium (RU) is supplied to the HWRs in sufficient quantities to support their operation both as power generators and americium burners. Two simplifying assumptions have been made. First, the economic value of RU is a linear function of the cost of fresh natural uranium (NU), and secondly, plutonium recycling for a third utility running a mixed oxide (MOX) fuelled reactor fleet has been already taking place, so that the extra cost of americium recycling is manageable. We conclude that, in order for this scenario to be economically attractive to the LWR operator, the averted disposal cost due to partitioning americium from LWR spent fuel must exceed 214 dollars per kg, comparable to estimates of the permanent disposal cost of the high level waste (HLW) from reprocessing spent LWR fuel. (authors)

  19. Decommissioning of the secondary containment of the steam generating heavy water reactor at UKAEA-Winfrith

    SciTech Connect

    Miller, Keith; Cornell, Rowland; Parkinson, Steve; McIntyre, Kevin; Staples, Andy

    2007-07-01

    Available in abstract form only. Full text of publication follows: The Winfrith SGHWR was a prototype nuclear power plant operated for 23 years by the United Kingdom Atomic Energy Authority (UKAEA) until 1990 when it was shut down permanently. The current Stage 1 decommissioning contract is part of a multi-stage strategy. It involves the removal of all the ancillary plant and equipment in the secondary containment and non-containment areas ahead of a series of contracts for the decommissioning of the primary containment, the reactor core and demolition of the building and all remaining facilities. As an outcome of a competitive tending process, the Stage 1 decommissioning contract was awarded to NUKEM with operations commencing in April 2005. The decommissioning processes involved with these plant items will be described with some emphasis of the establishment of multiple work-fronts for the production, recovery, treatment and disposal of mainly tritium-contaminated waste arising from its contact with the direct cycle reactor coolant. The means of size reduction of a variety of large, heavy and complex items of plant made from a range of materials will also be described with some emphasis on the control of fumes during hot cutting operations and establishing effective containments within a larger secondary containment structure. Disposal of these wastes in a timely and cost-effective manner is a major challenge facing the decommissioning team and has required the development of a highly efficient means of packing the resultant materials into mainly one-third height ISO containers for disposal as LLW. Details of the quantities of LLW and exempt wastes handled during this process will be given with a commentary about the difficulty in segregating these two waste streams efficiently. (authors)

  20. Numerical Analysis on the Calandria Tubes in the Moderator of a Heavy Water Reactor Using OpenFOAM and Other Codes

    NASA Astrophysics Data System (ADS)

    Chang, Se-Myong; Kim, Hyoung Tae

    2014-06-01

    CANDU, a prototype of heavy water reactor is modeled for the moderator system with porous media buoyancy-effect heat-transfer turbulence model. OpenFOAM, a set of C++ classes and libraries developed under the object-oriented concept, is selected as the tool of numerical analysis. The result from this computational code is compared with experiments and other commercial code data through ANSYS-CFX and COMSOL Multi-physics. The three-dimensional code concerning buoyancy force, turbulence, and heat transfer is tested and shown to be successful for the analysis of thermo-hydraulic system of heavy water reactors.

  1. Decommissioning of the secondary containment of the steam generating heavy water reactor at UKAEA Winfrith

    SciTech Connect

    Miller, K.D.; Cornell, R.M.; Parkinson, S.J.; McIntyre, K.; Staples, A.

    2007-07-01

    The Winfrith SGHWR was a prototype nuclear power plant operated for 23 years by the United Kingdom Atomic Energy Authority (UKAEA) until 1990 when it was shut down permanently. The current Stage 1 decommissioning contract is part of a multi-stage strategy. It involves the removal of all the ancillary plant and equipment in the secondary containment and non-containment areas ahead of a series of contracts for the decommissioning of the primary containment, the reactor core and demolition of the building and ail remaining facilities. As an outcome of a competitive tending process, the Stage 1 decommissioning contract was awarded to NUKEM with operations commencing in April 2005. The decommissioning processes involved with these plant items will be described with some emphasis of the establishment of multiple work-fronts for the production, recovery, treatment and disposal of mainly tritium-contaminated waste arising from its contact with the direct cycle reactor coolant. The means of size reduction of a variety of large, heavy and complex items of plant made from a range of materials will also be described with some emphasis on the control of fumes during hot cutting operations and establishing effective containments within a larger secondary containment structure. Disposal of these wastes in a timely and cost-effective manner is a major challenge facing the decommissioning team and has required the development of a highly efficient means of packing the resultant materials into mainly one-third height IS0 containers for disposal as LLW. Details of the quantities of LLW and exempt wastes handled during this process will be given with a commentary about the difficulty in segregating these two waste streams efficiently. The paper sets out to demonstrate the considerable progress that has been made with these challenging decommissioning operations at the SGHWR plant and to highlight some of the techniques and processes that have contributed to the overall success of the

  2. Inspection Head Design for the In-Service Inspection of Fuel Channels of Pressurized Heavy Water Reactors

    SciTech Connect

    Haruray, Amit Kumar; Veerapur, R.D.; Puri, R.K.; Singh, Manjit

    2006-07-01

    This paper discusses the challenges associated with the mechanical design of Inspection Head for the in-service inspection (ISI) of fuel channels of Indian Pressurized Heavy Water Reactors (PHWRs). ISI is carried out during shut down period in the reactor. Non Destructive Examination (NDE) of fuel channels is a mandatory requirement to acquire knowledge about the structural condition. A typical 220 MWe Reactor-core consists of 306 horizontal fuel channel assemblies (tubular in shape). There are typical design challenges due to their horizontal nature, long length (each assembly is around 9 meters long), and high radiation. Because of combined effect of above mentioned factors, these fuel channels develop permanent downward sag during service. This sag has to be negotiated by the Inspection Head. The Inspection Head houses all the NDE sensors and is deployed in the fuel channel with the help of reactor fuelling machine. It is driven inside the fuel channel by a separate external drive-system, which is capable of linearly advancing, retracting as well as rotating it all-round to achieve full-volumetric inspection. The paper also discusses an important design feature in the Inspection Head, which helps in maintaining a fixed distance between NDE sensors and the internal surface (ID) of the fuel channel, to enable us to obtain reliable and consistent inspection results. This objective is achieved with the help of two specially designed leaf-spring loaded roller modules, which are assembled in the Inspection Head at its front and rear, with NDE Sensor Module sandwiched between them. Another very important design feature in the Inspection Head helps the Spring-Loaded Roller Modules in carrying out their intended function of maintaining fixed distance despite the weight of the long drive extension links attached at the rear of Inspection Head or deviations due to any other reason. There are multiple drive extension links attached at the rear of the Inspection Head as the

  3. Evaluation of anticipatory signal to steam generator pressure control program for 700 MWe Indian pressurized heavy water reactor

    SciTech Connect

    Pahari, S.; Hajela, S.; Rammohan, H. P.; Malhotra, P. K.; Ghadge, S. G.

    2012-07-01

    700 MWe Indian Pressurized Heavy Water Reactor (IPHWR) is horizontal channel type reactor with partial boiling at channel outlet. Due to boiling, it has a large volume of vapor present in the primary loops. It has two primary loops connected with the help of pressurizer surge line. The pressurizer has a large capacity and is partly filled by liquid and partly by vapor. Large vapor volume improves compressibility of the system. During turbine trip or load rejection, pressure builds up in Steam Generator (SG). This leads to pressurization of Primary Heat Transport System (PHTS). To control pressurization of SG and PHTS, around 70% of the steam generated in SG is dumped into the condenser by opening Condenser Steam Dump Valves (CSDVs) and rest of the steam is released to the atmosphere by opening Atmospheric Steam Discharge Valves (ASDVs) immediately after sensing the event. This is accomplished by adding anticipatory signal to the output of SG pressure controller. Anticipatory signal is proportional to the thermal power of reactor and the proportionality constant is set so that SG pressure controller's output jacks up to ASDV opening range when operating at 100% FP. To simulate this behavior for 700 MWe IPHWR, Primary and secondary heat transport system is modeled. SG pressure control and other process control program have also been modeled to capture overall plant dynamics. Analysis has been carried out with 3-D neutron kinetics coupled thermal hydraulic computer code ATMIKA.T to evaluate the effect of the anticipatory signal on PHT pressure and over all plant dynamics during turbine trip in 700 MWe IPHWR. This paper brings out the results of the analysis with and without considering anticipatory signal in SG pressure control program during turbine trip. (authors)

  4. Analytical Study of High Concentration PCB Paint at the Heavy Water Components Test Reactor

    SciTech Connect

    Lowry, N.J.

    1998-10-21

    This report provides results of an analytical study of high concentration PCB paint in a shutdown nuclear test reactor located at the US Department of Energy's Savannah River Site (SRS). The study was designed to obtain data relevant for an evaluation of potential hazards associated with the use of and exposure to such paints.

  5. Reactor water cleanup system

    DOEpatents

    Gluntz, Douglas M.; Taft, William E.

    1994-01-01

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

  6. 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.

  7. Tritiated water processing for fusion reactors

    SciTech Connect

    Sood, S.K.; Kalyanam, K.M.

    1995-03-01

    Tritiated water represents a source of occupational exposure and environmental emissions for fusion and fission reactors. Fusion reactors must operate within stringent radionuclide emission limits. A range of tritiated water concentrations can be generated in fusion reactors, mostly in the form of tritiated light water. In contrast, tritium removal plants have been built in Canada and France to remove tritium from heavy water moderated fission reactors. Various isotope separation processes have been developed to remove tritium from light and heavy water. Appropriate process selection depends, amongst other items, on whether tritium is to be removed from light or heavy water, and on whether the detritiated water is recycled back to a process system or is discharged to the environment. This paper primarily discusses water detritiation requirements in fusion reactors and outlines process options that are suitable for meeting these requirements. 11 refs., 4 tabs.

  8. Design of PI(λ)D(μ) controller for global power control of Pressurized Heavy Water Reactor.

    PubMed

    Bongulwar, M R; Patre, B M

    2017-07-01

    In this paper, a robust stabilizing controller design method is presented for global power control of a Pressurized Heavy Water Reactor (PHWR) under step-back condition scheme using a Fractional Order Proportional Integral Derivative (PI(λ)D(μ)) controller resulting into robust performance. The method is applicable to design a controller for One Non Integer Order Plus Time Delay (NIOPTD-I) plant which satisfies design specifications such as phase margin and gain crossover frequency. Stability boundary locus method is used in (Kp, Ki, Kd) parameter space for NIOPTD-I plants to obtain stability region. The robust performance is obtained by satisfying flat phase condition at gain crossover frequency where phase is almost constant for large span of frequencies. The simulation result of the proposed PI(λ)D(μ) controller shows active step-back control to the insertion of the rod with no undershoot and with the robust performance, hence safe to the plant for gain variations from 500% lower side to 1000% upper side. The PI(λ)D(μ) controller with a plant shows that 30% and 50% global power drop from initial 100% is achieved in a reasonable time without undershoot. Copyright © 2017 ISA. Published by Elsevier Ltd. All rights reserved.

  9. Development of an internally cooled annular fuel bundle for pressurized heavy water reactors

    SciTech Connect

    Hamilton, H.; Armstrong, J.; Kittmer, A.; Zhuchkova, A.; Xu, R.; Hyland, B.; King, M.; Nava-Dominguez, A.; Livingstone, S.; Bergeron, A.

    2013-07-01

    A number of preliminary studies have been conducted at Atomic Energy of Canada Limited to explore the potential of using internally cooled annular fuel (ICAF) in CANDU reactors including finite element thermo-mechanical modelling, reactor physics, thermal hydraulics, fabrication and mechanical design. The most compelling argument for this design compared to the conventional solid-rod design is the significant reduction in maximum fuel temperature for equivalent LERs (linear element ratings). This feature presents the potential for power up-rating or higher burnup and a decreased defect probability due to in-core power increases. The thermal-mechanical evaluation confirmed the significant reduction in maximum fuel temperatures for ICAF fuel compared to solid-rod fuel for equivalent LER. The maximum fuel temperature increase as a function of LER increase is also significantly less for ICAF fuel. As a result, the sheath stress induced by an equivalent power increase is approximately six times less for ICAF fuel than solid-rod fuel. This suggests that the power-increase thresholds to failure (due to stress-corrosion cracking) for ICAF fuel should be well above those for solid-rod fuel, providing improvement in operation flexibility and safety.

  10. Breeding of 233U in the thorium-uranium fuel cycle in VVER reactors using heavy water

    NASA Astrophysics Data System (ADS)

    Marshalkin, V. E.; Povyshev, V. M.

    2015-12-01

    A method is proposed for achieving optimal neutron kinetics and efficient isotope transmutation in the 233U-232Th oxide fuel of water-moderated reactors with variable water composition (D2O, H2O) that ensures breeding of the 233U and 235U isotopes. The method is comparatively simple to implement.

  11. Breeding of {sup 233}U in the thorium–uranium fuel cycle in VVER reactors using heavy water

    SciTech Connect

    Marshalkin, V. E. Povyshev, V. M.

    2015-12-15

    A method is proposed for achieving optimal neutron kinetics and efficient isotope transmutation in the {sup 233}U–{sup 232}Th oxide fuel of water-moderated reactors with variable water composition (D{sub 2}O, H{sub 2}O) that ensures breeding of the {sup 233}U and {sup 235}U isotopes. The method is comparatively simple to implement.

  12. Controllability of depth dose distribution for neutron capture therapy at the Heavy Water Neutron Irradiation Facility of Kyoto University Research Reactor.

    PubMed

    Sakurai, Yoshinori; Kobayashi, Tooru

    2002-10-01

    The updating construction of the Heavy Water Neutron Irradiation Facility of the Kyoto University Research Reactor has been performed from November 1995 to March 1996 mainly for the improvement in neutron capture therapy. On the performance, the neutron irradiation modes with the variable energy spectra from almost pure thermal to epi-thermal neutrons became available by the control of the heavy-water thickness in the spectrum shifter and by the open-and-close of the cadmium and boral thermal neutron filters. The depth distributions of thermal, epi-thermal and fast neutron fluxes were measured by activation method using gold and indium, and the depth distributions of gamma-ray absorbed dose rate were measured using thermo-luminescent dosimeter of beryllium oxide for the several irradiation modes. From these measured data, the controllability of the depth dose distribution using the spectrum shifter and the thermal neutron filters was confirmed.

  13. An analysis of workers' tritium concentration in urine samples as a function of time after intake at Korean pressurised heavy water reactors.

    PubMed

    Kim, Hee Geun; Kong, Tae Young

    2012-12-01

    In general, internal exposure from tritium at pressurised heavy water reactors (PHWRs) accounts for ∼20-40 % of the total radiation dose. Tritium usually reaches the equilibrium concentration after a few hours inside the body and is then excreted from the body with an effective half-life in the order of 10 d. In this study, tritium metabolism was reviewed using its excretion rate in urine samples of workers at Korean PHWRs. The tritium concentration in workers' urine samples was also measured as a function of time after intake. On the basis of the monitoring results, changes in the tritium concentration inside the body were then analysed.

  14. Light water reactor program

    SciTech Connect

    Franks, S.M.

    1994-12-31

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

  15. Production of heavy water

    DOEpatents

    Spencer, Larry S.; Brown, Sam W.; Phillips, Michael R.

    2017-06-06

    Disclosed are methods and apparatuses for producing heavy water. In one embodiment, a catalyst is treated with high purity air or a mixture of gaseous nitrogen and oxygen with gaseous deuterium all together flowing over the catalyst to produce the heavy water. In an alternate embodiment, the deuterium is combusted to form the heavy water. In an alternate embodiment, gaseous deuterium and gaseous oxygen is flowed into a fuel cell to produce the heavy water. In various embodiments, the deuterium may be produced by a thermal decomposition and distillation process that involves heating solid lithium deuteride to form liquid lithium deuteride and then extracting the gaseous deuterium from the liquid lithium deuteride.

  16. DIANA: A multi-phase, multi-component hydrodynamic model for the analysis of severe accidents in heavy water reactors with multiple-tube assemblies

    SciTech Connect

    Tentner, A.M.

    1994-03-01

    A detailed hydrodynamic fuel relocation model has been developed for the analysis of severe accidents in Heavy Water Reactors with multiple-tube Assemblies. This model describes the Fuel Disruption and Relocation inside a nuclear fuel assembly and is designated by the acronym DIANA. DIANA solves the transient hydrodynamic equations for all the moving materials in the core and treats all the relevant flow regimes. The numerical solution techniques and some of the physical models included in DIANA have been developed taking advantage of the extensive experience accumulated in the development and validation of the LEVITATE (1) fuel relocation model of SAS4A [2, 3]. The model is designed to handle the fuel and cladding relocation in both voided and partially voided channels. It is able to treat a wide range of thermal/ hydraulic/neutronic conditions and the presence of various flow regimes at different axial locations within the same hydrodynamic channel.

  17. Radiological dose assessment for the dismantlement and decommissioning option for the Heavy Water Components Test Reactor facility at the Savannah River Site, Aiken, South Carolina

    SciTech Connect

    Faillace, E.R.; Kamboj, S.; Yu, C.; Chen, S.Y.

    1997-10-01

    Potential maximum radiation dose rates for a 10,000-year horizon were calculated for the dismantlement and decommissioning option for the Heavy Water Components Test Reactor facility at the Savannah River Site. The residual radioactive material guidelines (RESRAD) computer code was used. The study will help determine if it is acceptable (in terms of DOE radiation dose limits) for activated and contaminated concrete to remain in the facility, along with embedded radioactive piping and radioactive equipment. Four cases were developed to evaluate potential doses; the cases vary with regard to the definitions of the sources. Case A considers the dose from the reactor biological shield; case B considers the dose from contaminated concrete rubble; case C considers the dose from contaminated concrete rubble, the reactor biological shield, and installed equipment; and case D considers the dose from contaminated cuttings brought to the surface following the perforation of a well through the contaminated zone in case C. Site-specific parameter values were used to estimate the radiation doses. The results indicate that neither the DOE dose limit of 100 mrem/yr nor the 15-mrem/yr dose constraint would be exceeded for any of the cases. The potential maximum dose rates for cases A, B, C, and D are 0.000028, 0.015, 0.018, and 0.17 mrem/yr, respectively. The drinking water pathway is the dominant contributor to the doses in cases A through C, and the external gamma pathway is the dominant contributor in case D. Carbon-14, uranium-234, uranium-238, and americium-241 are the principal radionuclides contributing to the doses in cases A through C. Cobalt-60, europium-152, and barium-133 are the important radionuclides in case D. A sensitivity analysis was performed to determine which parameters have the greatest impact on the estimated doses. 9 refs., 11 figs., 3 tabs.

  18. Preliminary results of calculations for heavy-water nuclear-power-plant reactors employing {sup 235}U, {sup 233}U, and {sup 232}Th as a fuel and meeting requirements of a nonproliferation of nuclear weapons

    SciTech Connect

    Ioffe, B. L.; Kochurov, B. P.

    2012-02-15

    A physical design is developed for a gas-cooled heavy-water nuclear reactor intended for a project of a nuclear power plant. As a fuel, the reactor would employ thorium with a small admixture of enriched uranium that contains not more than 20% of {sup 235}U. It operates in the open-cycle mode involving {sup 233}U production from thorium and its subsequent burnup. The reactor meets the conditions of a nonproliferation of nuclear weapons: the content of fissionable isotopes in uranium at all stages of the process, including the final one, is below the threshold for constructing an atomic bomb, the amount of product plutonium being extremely small.

  19. Preliminary results of calculations for heavy-water nuclear-power-plant reactors employing 235U, 233U, and 232Th as a fuel and meeting requirements of a nonproliferation of nuclear weapons

    NASA Astrophysics Data System (ADS)

    Ioffe, B. L.; Kochurov, B. P.

    2012-02-01

    A physical design is developed for a gas-cooled heavy-water nuclear reactor intended for a project of a nuclear power plant. As a fuel, the reactor would employ thorium with a small admixture of enriched uranium that contains not more than 20% of 235U. It operates in the open-cycle mode involving 233U production from thorium and its subsequent burnup. The reactor meets the conditions of a nonproliferation of nuclear weapons: the content of fissionable isotopes in uranium at all stages of the process, including the final one, is below the threshold for constructing an atomic bomb, the amount of product plutonium being extremely small.

  20. Two-dimensional DORT discrete ordinates X-Y geometry neutron flux calculations for the Halden Heavy Boiling Water Reactor core configurations

    SciTech Connect

    Slater, C.O.

    1990-07-01

    Results are reported for two-dimensional discrete ordinates, X-Y geometry calculations performed for seven Halden Heavy Boiling Water Reactor core configurations. The calculations were performed in support of an effort to reassess the neutron fluence received by the reactor vessel. Nickel foil measurement data indicated considerable underprediction of fluences by the previously used multigroup removal- diffusion method. Therefore, calculations by a more accurate method were deemed appropriate. For each core configuration, data are presented for (1) integral fluxes in the core and near the vessel wall, (2) neutron spectra at selected locations, (3) isoflux contours superimposed on the geometry models, (4) plots of the geometry models, and (5) input for the calculations. The initial calculations were performed with several mesh sizes. Comparisons of the results from these calculations indicated that the uncertainty in the calculated fluxes should be less than 10%. However, three-dimensional effects (such as axial asymmetry in the fuel loading) could contribute to much greater uncertainty in the calculated neutron fluxes. 7 refs., 22 figs., 11 tabs.

  1. WATER BOILER REACTOR

    DOEpatents

    King, L.D.P.

    1960-11-22

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

  2. Utilization of non-weapons-grade plutonium and highly enriched uranium with breeding of the 233U isotope in the VVER reactors using thorium and heavy water

    NASA Astrophysics Data System (ADS)

    Marshalkin, V. E.; Povyshev, V. M.

    2015-12-01

    A method for joint utilization of non-weapons-grade plutonium and highly enriched uranium in the thorium-uranium—plutonium oxide fuel of a water-moderated reactor with a varying water composition (D2O, H2O) is proposed. The method is characterized by efficient breeding of the 233U isotope and safe reactor operation and is comparatively simple to implement.

  3. Utilization of non-weapons-grade plutonium and highly enriched uranium with breeding of the {sup 233}U isotope in the VVER reactors using thorium and heavy water

    SciTech Connect

    Marshalkin, V. E. Povyshev, V. M.

    2015-12-15

    A method for joint utilization of non-weapons-grade plutonium and highly enriched uranium in the thorium–uranium—plutonium oxide fuel of a water-moderated reactor with a varying water composition (D{sub 2}O, H{sub 2}O) is proposed. The method is characterized by efficient breeding of the {sup 233}U isotope and safe reactor operation and is comparatively simple to implement.

  4. Optimisation of the hot conditioning of carbon steel surfaces of primary heat transport system of Pressurized Heavy Water Reactors using electrochemical impedance spectroscopy

    NASA Astrophysics Data System (ADS)

    Kiran Kumar, M.; Gaonkar, Krishna; Ghosh, Swati; Kain, Vivekanand; Bojinov, Martin; Saario, Timo

    2010-06-01

    Hot conditioning operation of the primary heat transport system is an important step prior to the commissioning of Pressurized Heavy Water Reactors. One of the major objectives of the operation is to develop a stable and protective magnetite layer on the inner surfaces of carbon steel piping. The correlation between stable magnetite film growth on carbon steel surfaces and the period of exposure to hot conditioning environment is generally established by a combination of weight change measurements and microscopic/morphological observations of the specimens periodically removed during the operation. In the present study, electrochemical impedance spectroscopy (EIS) at room temperature is demonstrated as an alternate, quantitative technique to arrive at an optimal duration of the exposure period. Specimens of carbon steel were exposed for 24, 35 and 48 h during hot conditioning of primary heat transport system of two Indian PHWRs. The composition and morphology of oxide films grown during exposure was characterized by X-ray diffraction and optical microscopy. Further, ex situ electrochemical impedance spectra of magnetite films formed after each exposure were measured, in 1 ppm Li + electrolyte at room temperature as a function of potential in a range of -0.8 to +0.3 VSCE. The defect density of the magnetite films formed after each exposure was estimated by Mott-Schottky analysis of capacitances extracted from the impedance spectra. Further the ionic resistance of the oxide was also extracted from the impedance spectra. Defect density was observed to decrease with increase in exposure time and to saturate after 35 h, indicating stabilisation of the barrier layer part of the magnetite film. The values of the ionic transport resistance start to increase after 35-40 h of exposure. The quantitative ability of EIS technique to assess the film quality demonstrates that it can be used as a supplementary tool to the thickness and morphological characterizations of samples

  5. Light water reactor health physics.

    PubMed

    Prince, Robert J; Bradley, Scott E

    2004-11-01

    In this article an overview of the historical development of light water reactor health physics programs is presented. Operational health physics programs have developed and matured as experience in operating and maintaining light water reactors has been gained. Initial programs grew quickly in both size and complexity with the number and size of nuclear units under construction and in operation. Operational health physics programs evolved to face various challenges confronted by the nuclear industry, increasing the effectiveness of radiological safety measures. Industry improvements in radiological safety performance have resulted in significant decreases in annual collective exposures from a high value of 790 person-rem in 1980 to 117 person-rem per reactor in 2002. Though significant gains have been made, the continued viability of the nuclear power industry is confronted with an aging workforce, as well as the challenges posed by deregulation and the need to maintain operational excellence.

  6. Light water reactor health physics.

    PubMed

    Prince, Robert J; Bradley, Scott E

    2005-06-01

    In this article an overview of the historical development of light water reactor health physics programs is presented. Operational health physics programs have developed and matured as experience in operating and maintaining light water reactors has been gained. Initial programs grew quickly in both size and complexity with the number and size of nuclear units under construction and in operation. Operational health physics programs evolved to face various challenges confronted by the nuclear industry, increasing the effectiveness of radiological safety measures. Industry improvements in radiological safety performance have resulted in significant decreases in annual collective exposures from a high value of 790 person-rem in 1980 to 117 person-rem per reactor in 2002. Though significant gains have been made, the continued viability of the nuclear power industry is confronted with an aging workforce, as well as the challenges posed by deregulation and the need to maintain operational excellence.

  7. CHIMNEY FOR BOILING WATER REACTOR

    DOEpatents

    Petrick, M.

    1961-08-01

    A boiling-water reactor is described which has vertical fuel-containing channels for forming steam from water. Risers above the channels increase the head of water radially outward, whereby water is moved upward through the channels with greater force. The risers are concentric and the radial width of the space between them is somewhat small. There is a relatively low rate of flow of water up through the radially outer fuel-containing channels, with which the space between the risers is in communication. (AE C)

  8. Removal of gadolinium nitrate from heavy water

    SciTech Connect

    Wilde, E.W.

    2000-03-22

    Work was conducted to develop a cost-effective process to purify 181 55-gallon drums containing spent heavy water moderator (D2O) contaminated with high concentrations of gadolinium nitrate, a chemical used as a neutron poison during former nuclear reactor operations at the Savannah River Site (SRS). These drums also contain low level radioactive contamination, including tritium, which complicates treatment options. Presently, the drums of degraded moderator are being stored on site. It was suggested that a process utilizing biological mechanisms could potentially lower the total cost of heavy water purification by allowing the use of smaller equipment with less product loss and a reduction in the quantity of secondary waste materials produced by the current baseline process (ion exchange).

  9. Water-cooled nuclear reactor

    SciTech Connect

    Braun, W.; Irion, L.

    1981-07-07

    A water-cooled nuclear reactor pressure vessel is internally provided with a check valve assembly for the inner end of a coolant inlet nozzle to prevent immediate loss of the coolant from the vessel by backward flow in the event the pressure of the coolant supply through the nozzle, drops below the pressure inside the vessel while the reactor is operating. The valve assembly includes a plurality of flap valves working in parallel so that if one fails to close when required, the balance of the flap valves remain operative to check the discharge of the coolant.

  10. LIGHT WATER MODERATED NEUTRONIC REACTOR

    DOEpatents

    Christy, R.F.; Weinberg, A.M.

    1957-09-17

    A uranium fuel reactor designed to utilize light water as a moderator is described. The reactor core is in a tank at the bottom of a substantially cylindrical cross-section pit, the core being supported by an apertured grid member and comprised of hexagonal tubes each containing a pluralily of fuel rods held in a geometrical arrangement between end caps of the tubes. The end caps are apertured to permit passage of the coolant water through the tubes and the fuel elements are aluminum clad to prevent corrosion. The tubes are hexagonally arranged in the center of the tank providing an amulus between the core and tank wall which is filled with water to serve as a reflector. In use, the entire pit and tank are filled with water in which is circulated during operation by coming in at the bottom of the tank, passing upwardly through the grid member and fuel tubes and carried off near the top of the pit, thereby picking up the heat generated by the fuel elements during the fission thereof. With this particular design the light water coolant can also be used as the moderator when the uranium is enriched by fissionable isotope to an abundance of U/sup 235/ between 0.78% and 2%.

  11. SUPERHEATING IN A BOILING WATER REACTOR

    DOEpatents

    Treshow, M.

    1960-05-31

    A boiling-water reactor is described in which the steam developed in the reactor is superheated in the reactor. This is accomplished by providing means for separating the steam from the water and passing the steam over a surface of the fissionable material which is not in contact with the water. Specifically water is boiled on the outside of tubular fuel elements and the steam is superheated on the inside of the fuel elements.

  12. Feasibility study on the thorium fueled boiling water breeder reactor

    SciTech Connect

    PetrusTakaki, N.

    2012-07-01

    The feasibility of (Th,U)O 2 fueled, boiling water breeder reactor based on conventional BWR technology has been studied. In order to determine the potential use of water cooled thorium reactor as a competitive breeder, this study evaluated criticality, breeding and void reactivity coefficient in response to changes made in MFR and fissile enrichments. The result of the study shows that while using light water as moderator, low moderator to fuel volume ratio (MFR=0.5), it was possible to breed fissile fuel in negative void reactivity condition. However the burnup value was lower than the value of the current LWR. On the other hand, heavy water cooled reactor shows relatively wider feasible breeding region, which lead into possibility of designing a core having better neutronic and economic performance than light water with negative void reactivity coefficient. (authors)

  13. Review of High Temperature Water and Steam Cooled Reactor Concepts

    SciTech Connect

    Oka, Yoshiaki

    2002-07-01

    This review summarizes design concepts of supercritical-pressure water cooled reactors (SCR), nuclear superheaters and steam cooled fast reactors from 1950's to the present time. It includes water moderated supercritical steam cooled reactor, SCOTT-R and SC-PWR of Westinghouse, heavy water moderated light water cooled SCR of GE, SCLWR and SCFR of the University of Tokyo, B-500SKDI of Kurchatov Institute, CANDU -X of AECL, nuclear superheaters of GE, subcritical-pressure steam cooled FBR of KFK and B and W, Supercritical-pressure steam cooled FBR of B and W, subcritical-pressure steam cooled high converter by Edlund and Schultz and subcritical-pressure water-steam cooled FBR by Alekseev. This paper is prepared based on the previous review of SCR2000 symposium, and some author's comments are added. (author)

  14. PRESSURIZED WATER REACTOR CORE WITH PLUTONIUM BURNUP

    DOEpatents

    Puechl, K.H.

    1963-09-24

    A pressurized water reactor is described having a core containing Pu/sup 240/ in which the effective microscopic neutronabsorption cross section of Pu/sup 240/ in unconverted condition decreases as the time of operation of the reactor increases, in order to compensate for loss of reactivity resulting from fission product buildup during reactor operation. This means serves to improve the efficiency of the reactor operation by reducing power losses resulting from control rods and burnable poisons. (AEC)

  15. Light-Water Breeder Reactor

    DOEpatents

    Beaudoin, B. R.; Cohen, J. D.; Jones, D. H.; Marier, Jr, L. J.; Raab, H. F.

    1972-06-20

    Described is a light-water-moderated and -cooled nuclear breeder reactor of the seed-blanket type characterized by core modules comprising loosely packed blanket zones enriched with fissile fuel and axial zoning in the seed and blanket regions within each core module. Reactivity control over lifetime is achieved by axial displacement of movable seed zones without the use of poison rods in the embodiment illustrated. The seed is further characterized by a hydrogen-to-uranium-233 atom ratio in the range 10 to 200 and a uranium-233-to-thorium-232 atom ratio ranging from 0.012 to 0.200. The seed occupies from 10 to 35 percent of the core volume in the form of one or more individual islands or annuli. (NSA 26: 55130)

  16. Physical particularities of nuclear reactors using heavy moderators of neutrons

    SciTech Connect

    Kulikov, G. G. Shmelev, A. N.

    2016-12-15

    In nuclear reactors, thermal neutron spectra are formed using moderators with small atomic weights. For fast reactors, inserting such moderators in the core may create problems since they efficiently decelerate the neutrons. In order to form an intermediate neutron spectrum, it is preferable to employ neutron moderators with sufficiently large atomic weights, using {sup 233}U as a fissile nuclide and {sup 232}Th and {sup 231}Pa as fertile ones. The aim of the work is to investigate the properties of heavy neutron moderators and to assess their advantages. The analysis employs the JENDL-4.0 nuclear data library and the SCALE program package for simulating the variation of fuel composition caused by irradiation in the reactor. The following main results are obtained. By using heavy moderators with small neutron moderation steps, one is able to (1) increase the rate of resonance capture, so that the amount of fertile material in the fuel may be reduced while maintaining the breeding factor of the core; (2) use the vacant space for improving the fuel-element properties by adding inert, strong, and thermally conductive materials and by implementing dispersive fuel elements in which the fissile material is self-replenished and neutron multiplication remains stable during the process of fuel burnup; and (3) employ mixtures of different fertile materials with resonance capture cross sections in order to increase the resonance-lattice density and the probability of resonance neutron capture leading to formation of fissile material. The general conclusion is that, by forming an intermediate neutron spectrum with heavy neutron moderators, one can use the fuel more efficiently and improve nuclear safety.

  17. Physical particularities of nuclear reactors using heavy moderators of neutrons

    NASA Astrophysics Data System (ADS)

    Kulikov, G. G.; Shmelev, A. N.

    2016-12-01

    In nuclear reactors, thermal neutron spectra are formed using moderators with small atomic weights. For fast reactors, inserting such moderators in the core may create problems since they efficiently decelerate the neutrons. In order to form an intermediate neutron spectrum, it is preferable to employ neutron moderators with sufficiently large atomic weights, using 233U as a fissile nuclide and 232Th and 231Pa as fertile ones. The aim of the work is to investigate the properties of heavy neutron moderators and to assess their advantages. The analysis employs the JENDL-4.0 nuclear data library and the SCALE program package for simulating the variation of fuel composition caused by irradiation in the reactor. The following main results are obtained. By using heavy moderators with small neutron moderation steps, one is able to (1) increase the rate of resonance capture, so that the amount of fertile material in the fuel may be reduced while maintaining the breeding factor of the core; (2) use the vacant space for improving the fuel-element properties by adding inert, strong, and thermally conductive materials and by implementing dispersive fuel elements in which the fissile material is self-replenished and neutron multiplication remains stable during the process of fuel burnup; and (3) employ mixtures of different fertile materials with resonance capture cross sections in order to increase the resonance-lattice density and the probability of resonance neutron capture leading to formation of fissile material. The general conclusion is that, by forming an intermediate neutron spectrum with heavy neutron moderators, one can use the fuel more efficiently and improve nuclear safety.

  18. Missiles caused by severe pressurized-water reactor accidients

    SciTech Connect

    Krieg, R.

    1995-07-01

    For future pressurized-water reactors, which should be designed against core-meltdown accidents, missiles generated inside the containment present a severe problem for its integrity. The masses and geometries of the missiles, as well as their velocities, may vary to a great extent. Therefore a reliable proof of the containment integrity is very difficult. In this article the potential sources of missiles are discussed, and the conclusion was reached that the generation of heavy missiles must be prevented. Steam explosions must not damage the reactor vessel head. Thus fragments of the head cannot become missiles that endanger the containment shell. Furthermore, during a melt-through failure of the reactor vessel under high pressure, the resulting forces must not catapult the whole vessel against the containment shell. Only missiles caused by hydrogen explosions may be tolerable, but shielding structures that protect the containment shell may be required. Further investigations are necessary. Finally, measures are described showing that the generation of heavy missiles can indeed be prevented. Investigations are currently being carried out that will confirm the strength of the reactor vessel head. In addition, a device for retaining the fragments of a failing reactor vessel is discussed.

  19. (Severe accident technology of BWR (Boiling Water Reactor) reactors)

    SciTech Connect

    Ott, L.J.

    1989-10-23

    The traveler attended the 1989 CORA Workshop at KfK, FRG. Participation included the presentation included the presentation of three papers on work performed by the Boiling Water Reactor Severe Accident Technology (BWRSAT) program at Oak Ridge National Laboratory (ORNL) in Boiling Water Reactor (BWR) severe accident analyses. The Statement of Work (June 1989) for the BWRSAT Program provides for code analyses of the BWR CORA experiments performed at KfK. Additionally, it is intended that BWRSAT personnel participate in the planning process for future CORA BWR experiments. For these purposes, meetings were held with KfK staff to arrange for acquisition of detailed CORA facility drawings, experimental data, and related engineering. 17 refs.

  20. Pressurized water reactor flow skirt apparatus

    SciTech Connect

    Kielb, John F.; Schwirian, Richard E.; Lee, Naugab E.; Forsyth, David R.

    2016-04-05

    A pressurized water reactor vessel having a flow skirt formed from a perforated cylinder structure supported in the lower reactor vessel head at the outlet of the downcomer annulus, that channels the coolant flow through flow holes in the wall of the cylinder structure. The flow skirt is supported at a plurality of circumferentially spaced locations on the lower reactor vessel head that are not equally spaced or vertically aligned with the core barrel attachment points, and the flow skirt employs a unique arrangement of hole patterns that assure a substantially balanced pressure and flow of the coolant over the entire underside of the lower core support plate.

  1. Heavy reflector experiments in the IPEN/MB-01 reactor: Stainless steel, carbon steel and nickel

    SciTech Connect

    Santos, Adimir dos; Andrade e Silva, Graciete Simoes de; Jerez, Rogerio; Liambos Mura, Luis Felipe; Fuga, Rinaldo

    2013-05-06

    New experiments devoted to the measurements of physical parameters of a light water core surrounded by a heavy reflector were performed in the IPEN/MB-01 research reactor facility. These experiments comprise three sets of heavy reflector (SS-304, Carbon Steel, and Nickel) in a form of laminates around 3 mm thick. Each set was introduced individually in the west face of the core of the IPEN/MB-01 reactor. The aim here is to provide high quality experimental data for the interpretation and validation of the SS-304 heavy reflector calculation methods. The experiments of Carbon Steel, which is composed mainly of iron, and Nickel were performed to provide a consistent and an interpretative check for the SS-304 reflector experiment. The experimental results comprise critical control bank positions, temperatures and reactivities as a function of the number of the plates. Particularly to the case of Nickel, the experimental data are unique of its kind. The theoretical analysis was performed by MCNP-5 with the nuclear data library ENDF/B-VII.0. It was shown that this nuclear data library has a very good performance up to thirteen plates and overestimates the reactivity for higher number of plates independently of the type of the reflector.

  2. TA-2 water boiler reactor decommissioning (Phase 1)

    SciTech Connect

    Elder, J.C.; Knoell, C.L.

    1986-12-01

    Removal of external structures and underground piping associated with the gaseous effluent (stack) line from the TA-2 Water Boiler Reactor was performed as Phase I of reactor decommissioning. Six concrete structures were dismantled and 435 ft of contaminated underground piping was removed. Extensive soil contamination by /sup 137/Cs was encountered around structure TA-2-48 and in a suspected leach field near the stream flowing through Los Alamos Canyon. Efforts to remove all contaminated soil were hampered by infiltrating ground water and heavy rains. Methods, cleanup guidelines, and ALARA decisions used to successfully restore the area are described. The cost of the project was approximately $320K; 970 m/sup 3/ of low-level solid radioactive waste resulted from the cleanup operations.

  3. Tritium issues in commercial pressurized water reactors

    SciTech Connect

    Jones, G.

    2008-07-15

    Tritium has become an important radionuclide in commercial Pressurized Water Reactors because of its mobility and tendency to concentrate in plant systems as tritiated water during the recycling of reactor coolant. Small quantities of tritium are released in routine regulated effluents as liquid water and as water vapor. Tritium has become a focus of attention at commercial nuclear power plants in recent years due to inadvertent, low-level, chronic releases arising from routine maintenance operations and from component failures. Tritium has been observed in groundwater in the vicinity of stations. The nuclear industry has undertaken strong proactive corrective measures to prevent recurrence, and continues to eliminate emission sources through its singular focus on public safety and environmental stewardship. This paper will discuss: production mechanisms for tritium, transport mechanisms from the reactor through plant, systems to the environment, examples of routine effluent releases, offsite doses, basic groundwater transport and geological issues, and recent nuclear industry environmental and legal ramifications. (authors)

  4. Analysis of the TRIGA Reactor Pool Water

    DTIC Science & Technology

    1993-08-01

    AD-A270 956 L11L1I~I1 11 11 :1Ji ili! August 1993 AFRRI 93-5 TECHNICAL REPORT Analysis of the TRIGA Reactor Pool Water L OCT 1 93 John Dickson Robert...COVER~ED I August 1993 Technical Report 4 TITLE AND SUBTITLE S.FNDN NUMBERS Analysis of the TRIGA Reactor Pool Water PE: NWED QAXM 6, AUTHOR(S) Dickson...AVAILABIIY STATEMENT 1 2b. DISTRIBUTION CODE Approved for public release; distribution unlimited. 13. ABSTRACT tMaxtm -um 200 words ) 14. SUBJECTTERMS 1S

  5. SELF-REGULATING BOILING-WATER NUCLEAR REACTORS

    DOEpatents

    Ransohoff, J.A.; Plawchan, J.D.

    1960-08-16

    A boiling-water reactor was designed which comprises a pressure vessel containing a mass of water, a reactor core submerged within the water, a reflector tank disposed within the reactor, the reflector tank being open at the top to the interior of the pressure vessel, and a surge tank connected to the reflector tank. In operation the reflector level changes as a function of the pressure witoin the reactor so that the reactivity of the reactor is automatically controlled.

  6. Optimized heavy ion beam probing for International Thermonuclear Experimental Reactor

    NASA Astrophysics Data System (ADS)

    Melnikov, A. V.; Eliseev, L. G.

    1999-01-01

    The international workgroup developed the conceptual design of a heavy ion beam probe (HIBP) diagnostics for International Thermonuclear Experimental Reactor (ITER), which is intended for measurements of the plasma potential profile in a gradient area. Now we optimized it by the accurate analysis of the probing trajectories and variation of positions of the injection and detection points. Optimization allows us to reduce the energy of Tl+ beam from 5.6 to 3.4 MeV for standard ITER regime. The detector line starting at the plasma edge towards the center can get an outer part of the horizontal radial potential profile by variation of the energy. The observed radial interval is slightly increased up to 0.76<ρ<1 with respect to initial version 0.8<ρ<1, that allows to cover the region of the density gradient more reliably. Almost double reduction of the beam energy is a critical point. Thus we can significantly decrease the sizes of the accelerator and energy analyzer, the cost of the equipment, and impact of the diagnostics to the machine. Therefore the optimized HIBP design can be realized in ITER.

  7. Water issues associated with heavy oil production.

    SciTech Connect

    Veil, J. A.; Quinn, J. J.; Environmental Science Division

    2008-11-28

    Crude oil occurs in many different forms throughout the world. An important characteristic of crude oil that affects the ease with which it can be produced is its density and viscosity. Lighter crude oil typically can be produced more easily and at lower cost than heavier crude oil. Historically, much of the nation's oil supply came from domestic or international light or medium crude oil sources. California's extensive heavy oil production for more than a century is a notable exception. Oil and gas companies are actively looking toward heavier crude oil sources to help meet demands and to take advantage of large heavy oil reserves located in North and South America. Heavy oil includes very viscous oil resources like those found in some fields in California and Venezuela, oil shale, and tar sands (called oil sands in Canada). These are described in more detail in the next chapter. Water is integrally associated with conventional oil production. Produced water is the largest byproduct associated with oil production. The cost of managing large volumes of produced water is an important component of the overall cost of producing oil. Most mature oil fields rely on injected water to maintain formation pressure during production. The processes involved with heavy oil production often require external water supplies for steam generation, washing, and other steps. While some heavy oil processes generate produced water, others generate different types of industrial wastewater. Management and disposition of the wastewater presents challenges and costs for the operators. This report describes water requirements relating to heavy oil production and potential sources for that water. The report also describes how water is used and the resulting water quality impacts associated with heavy oil production.

  8. Hydrogen and water reactor safety: proceedings

    SciTech Connect

    Not Available

    1982-01-01

    Separate abstracts were prepared for papers presented in the following areas of interest: 1) hydrogen research programs; 2) hydrogen behavior during light water reactor accidents; 3) combustible gas generation; 4) hydrogen transport and mixing; 5) combustion modeling and experiments; 6) accelerated flames and detonations; 7) combustion mitigation and control; and 8) equipment survivability.

  9. Towards intrinsically safe light-water reactors

    SciTech Connect

    Hannerz, K

    1983-07-01

    Most of the present impediments to the rational use of the nuclear option have their roots in the reactor safety issue. The approach taken to satisfy the escalating safety concerns has resulted in excessively complex and expensive plant designs but has failed to create public confidence. This paper describes a new approach based on the principle of Process Inherent Ultimate Safety (PIUS). With the PIUS principle, ultimate safety is obtained by guaranteeing core integrity under all credible conditions. This is accomplished on the basis of the laws of gravity and thermohydraulics alone, interacting with the heat extraction process in an intact or damaged primary circuit, without recourse to engineered safety systems that may fail or dependence on error-prone human intervention. Application of the PIUS principle to the pressurized water reactor involves a substantial redesign of the reactor and primary system but builds on established PWR technology where long-term operation is needed for verification.

  10. BOILING WATER REACTOR WITH FEED WATER INJECTION NOZZLES

    DOEpatents

    Treshow, M.

    1963-04-30

    This patent covers the use of injection nozzles for pumping water into the lower ends of reactor fuel tubes in which water is converted directly to steam. Pumping water through fuel tubes of this type of boiling water reactor increases its power. The injection nozzles decrease the size of pump needed, because the pump handles only the water going through the nozzles, additional water being sucked into the tubes by the nozzles independently of the pump from the exterior body of water in which the fuel tubes are immersed. The resulting movement of exterior water along the tubes holds down steam formation, and thus maintains the moderator effectiveness, of the exterior body of water. (AEC)

  11. Electrochemistry of Water-Cooled Nuclear Reactors

    SciTech Connect

    Macdonald, Dgiby; Urquidi-Macdonald, Mirna; Pitt, Jonathan

    2006-08-08

    This project developed a comprehensive mathematical and simulation model for calculating thermal hydraulic, electrochemical, and corrosion parameters, viz. temperature, fluid flow velocity, pH, corrosion potential, hydrogen injection, oxygen contamination, stress corrosion cracking, crack growth rate, and other important quantities in the coolant circuits of water-cooled nuclear power plants, including both Boiling Water Reactors (BWRs) and Pressurized Water Reactors (PWRs). The model is being used to assess the three major operational problems in Pressurized Water Reactors (PWR), which include mass transport, activity transport, and the axial offset anomaly, and provide a powerful tool for predicting the accumulation of SCC damage in BWR primary coolant circuits as a function of operating history. Another achievement of the project is the development of a simulation tool to serve both as a training tool for plant operators and as an engineering test-bed to evaluate new equipment and operating strategies (normal operation, cold shut down and others). The development and implementation of the model allows us to estimate the activity transport or "radiation fields" around the primary loop and the vessel, as a function of the operating parameters and the water chemistry.

  12. PIPING FOR COOLANT WATER IS INSTALLED INSIDE REACTOR STRUCTURE PRIOR ...

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

    PIPING FOR COOLANT WATER IS INSTALLED INSIDE REACTOR STRUCTURE PRIOR TO EMBEDMENT IN CONCRETE. HIGHER PIPE IS INLET; THE OTHER, THE OUTLET LOOP. INLET PIPE WILL CONNECT TO TOP SECTION OF REACTOR VESSEL. INL NEGATIVE NO. 1287. Unknown Photographer, 1/18/1951 - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

  13. NEUTRONIC REACTORS

    DOEpatents

    Wigner, E.P.

    1960-11-22

    A nuclear reactor is described wherein horizontal rods of thermal- neutron-fissionable material are disposed in a body of heavy water and extend through and are supported by spaced parallel walls of graphite.

  14. ``Heavy-water Lattice and Heavy-Quark''

    NASA Astrophysics Data System (ADS)

    Maksoed, Ssi, Wh-

    Refer to Birgitt Roettger-Roessler: ``Feelings at the Margins'', 2014 retrieved the Vienna, 2006 UNIDO Research Programme: Combating Marginalization and Poverty through Industrial Development/COMPID. Also from Vienna, on Feb 18-22, 1963 reported Technical Report Series 20 about ``Heavy Water Lattice''. Failed to relates scale-invariant properties of public-Debt growth to convergence in perturbation theory, sought JH Field: ``Convergence & Gauge Dependence Properties:..''. Furthers, in GP Lepage: ``On the Viabilities of Lattice Perturbation Theory'', 1992 stated: ``in terms of physical quantities, like the heavy-quark potential, greatly enhanced the predictive power of lattice perturbation theory''. Acknowledgements to HE. Mr. H. TUK SETYOHADI, Jl. Sriwijaya Raya 3, South-Jakarta, INDONESIA.

  15. Selective removal of heavy metals from metal-bearing wastewater in a cascade line reactor.

    PubMed

    Pavlović, Jelena; Stopić, Srećko; Friedrich, Bernd; Kamberović, Zeljko

    2007-11-01

    This paper is a part of the research work on 'Integrated treatment of industrial wastes towards prevention of regional water resources contamination - INTREAT' the project. It addresses the environmental pollution problems associated with solid and liquid waste/effluents produced by sulfide ore mining and metallurgical activities in the Copper Mining and Smelting Complex Bor (RTB-BOR), Serbia. However, since the minimum solubility for the different metals usually found in the polluted water occurs at different pH values and the hydroxide precipitates are amphoteric in nature, selective removal of mixed metals could be achieved as the multiple stage precipitation. For this reason, acid mine water had to be treated in multiple stages in a continuous precipitation system-cascade line reactor. All experiments were performed using synthetic metal-bearing effluent with chemical a composition similar to the effluent from open pit, Copper Mining and Smelting Complex Bor (RTB-BOR). That effluent is characterized by low pH (1.78) due to the content of sulfuric acid and heavy metals, such as Cu, Fe, Ni, Mn, Zn with concentrations of 76.680, 26.130, 0.113, 11.490, 1.020 mg/dm3, respectively. The cascade line reactor is equipped with the following components: for feeding of effluents, for injection of the precipitation agent, for pH measurements and control, and for removal of the process gases. The precipitation agent was 1M NaOH. In each of the three reactors, a changing of pH and temperature was observed. In order to verify. efficiency of heavy metals removal, chemical analyses of samples taken at different pH was done using AES-ICP. Consumption of NaOH in reactors was 370 cm3, 40 cm3 and 80 cm3, respectively. Total time of the experiment was 4 h including feeding of the first reactor. The time necessary to achieve the defined pH value was 25 min for the first reactor and 13 min for both second and third reactors. Taking into account the complete process in the cascade line

  16. Start-up procedures and analysis of heavy metals inhibition on methanogenic activity in EGSB reactor.

    PubMed

    Colussi, I; Cortesi, A; Della Vedova, L; Gallo, V; Robles, F K Cano

    2009-12-01

    The effectiveness of operating an industrial UASB reactor, treating wastewater from the beer industry, with flows containing heavy metals was evaluated. A pilot-scale UASB reactor, already used to simulate the industrial reactor, was unsuccessfully employed. An easy start-up was obtained arranging it as an EGSB reactor. Considerations about this modification are reported. The effects of Cu(II), Ni(II) and Cr(III) ions on the anaerobic activity were analyzed by measurements of methane production rate and COD removal. The employed biomass was the sludge of the industrial UASB reactor, while a solution of ethanol and sodium acetate with COD of 3000 mg/L and a heavy metal concentration of 50 mg/L were continuously fed. Experimental results proved higher biomass sensitivity for copper and much slighter for nickel and chromium. Moreover, copper inhibition has been demonstrated to be less significant if a metal-free feed was provided to the system before copper addition.

  17. 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.

  18. Containment system for supercritical water oxidation reactor

    DOEpatents

    Chastagner, Philippe

    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.

  19. Thermophysical properties of saturated light and heavy water for Advanced Neutron Source applications

    SciTech Connect

    Crabtree, A.; Siman-Tov, M.

    1993-05-01

    The Advanced Neutron Source is an experimental facility being developed by Oak Ridge National Laboratory. As a new nuclear fission research reactor of unprecedented flux, the Advanced Neutron Source Reactor will provide the most intense steady-state beams of neutrons in the world. The high heat fluxes generated in the reactor [303 MW(t) with an average power density of 4.5 MW/L] will be accommodated by a flow of heavy water through the core at high velocities. In support of this experimental and analytical effort, a reliable, highly accurate, and uniform source of thermodynamic and transport property correlations for saturated light and heavy water were developed. In order to attain high accuracy in the correlations, the range of these correlations was limited to the proposed Advanced Neutron Source Reactor`s nominal operating conditions. The temperature and corresponding saturation pressure ranges used for light water were 20--300{degrees}C and 0.0025--8.5 MPa, respectively, while those for heavy water were 50--250{degrees}C and 0.012--3.9 MPa. Deviations between the correlation predictions and data from the various sources did not exceed 1.0%. Light water vapor density was the only exception, with an error of 1.76%. The physical property package consists of analytical correlations, SAS codes, and FORTRAN subroutines incorporating these correlations, as well as an interactive, easy-to-use program entitled QuikProp.

  20. Modifications of water chemistry for pressurized water reactors

    SciTech Connect

    Fletcher, W.D.

    1987-01-01

    For commercial pressurized water reactors, the evolution of the water chemistry for both the reactor coolant systems (RCS) and the balance of plant (BOP) systems has been in concert with or responsive to the changes in specified materials and operating experience. Much of the early reactor coolant chemistry was identified within a program to develop the use of boric acid as a water soluble neutron absorber for chemical shim. Boric acid was selected as the absorber. The BOP steam and feedwater cycle chemistry was largely a translation from conventional low-pressure boiler water chemistry with specific changes to account for new materials in major components. Today's chemistry for the reactor and BOP systems has undergone significant changes since circa 1950. With advances in the design duty of the plant components and with a much better understanding of the mechanisms of corrosion of component materials, the water chemistry practices and methods of control are more successfully aimed at plant reliability than ever before. This paper presents a review of several important advances in water chemistry, as applied to both the RCS and the BOP and the basis for their adoption.

  1. Boiling water neutronic reactor incorporating a process inherent safety design

    DOEpatents

    Forsberg, C.W.

    1985-02-19

    A boiling-water reactor core is positioned within a prestressed concrete reactor vessel of a size which will hold a supply of coolant water sufficient to submerge and cool the reactor core by boiling for a period of at least one week after shutdown. Separate volumes of hot, clean (nonborated) water for cooling during normal operation and cool highly borated water for emergency cooling and reactor shutdown are separated by an insulated wall during normal reactor operation with contact between the two water volumes being maintained at interfaces near the top and bottom ends of the reactor vessel. Means are provided for balancing the pressure of the two water volumes at the lower interface zone during normal operation to prevent entry of the cool borated water into the reactor core region, for detecting the onset of excessive power to coolant flow conditions in the reactor core and for detecting low water levels of reactor coolant. Cool borated water is permitted to flow into the reactor core when low reactor coolant levels or excessive power to coolant flow conditions are encountered.

  2. SWR 1000: The Innovative Boiling Water Reactor

    SciTech Connect

    Brettschuh, Werner; Hudson, Greg

    2004-07-01

    Framatome ANP has developed the boiling water reactor SWR 1000 in close cooperation with German nuclear utilities and with support from various European partners. This advanced reactor design marks a new era in the successful tradition of boiling water reactor technology and, with a gross electric output of between 1290 and 1330 MW, is aimed at assuring competitive power generating costs compared to gas- and coal-fired stations. At the same time, the SWR 1000 meets the highest safety standards, including control of a core melt accident these objectives are met by supplementing active safety systems with passive safety equipment of diverse design for accident detection and control and by simplifying systems needed for normal plant operation on the basis of past operating experience. The plant is also protected against airplane crash loads. A short construction period, flexible fuel cycle lengths of between 12 and 24 months and a high fuel discharge burn-up all contribute towards meeting economic goals. The SWR 1000 fulfills international nuclear regulatory requirements and has been offered to TVO for the fifth nuclear unit in Finland. (authors)

  3. Comparison of heavy metal toxicity in continuous flow and batch reactors

    NASA Astrophysics Data System (ADS)

    Sengor, S. S.; Gikas, P.; Moberly, J. G.; Peyton, B. M.; Ginn, T. R.

    2009-12-01

    The presence of heavy metals may significantly affect microbial growth. In many cases, small amounts of particular heavy metals may stimulate microbial growth; however, larger quantities may result in microbial growth reduction. Environmental parameters, such as growth pattern may alter the critical heavy metal concentration, above which microbial growth stimulation turns to growth inhibition. Thus, it is important to quantify the effects of heavy metals on microbial activity for understanding natural or manmade biological reactors, either in situ or ex situ. Here we compare the toxicity of Zn and Cu on Arthrobacter sp., a heavy metal tolerant microorganism, under continuous flow versus batch reactor operations. Batch and continuous growth tests of Arthrobacter sp. were carried out at various individual and combined concentrations of Zn and Cu. Biomass concentration (OD) was measured for both the batch and continuous reactors, whereas ATP, oxygen uptake rates and substrate concentrations were additionally measured for the continuous system. Results indicated that Cu was more toxic than Zn under all conditions for both systems. In batch reactors, all tested Zn concentrations up to 150 uM showed a stimulatory effect on microbial growth. However, in the case of mixed Zn and Cu exposures, the presence of Zn either eliminated (at the 50 uM level both Zn and Cu) or reduced by ~25% (at the 100 and 150 uM levels both Zn and Cu) the Cu-induced inhibition. In the continuous system, only one test involved combined Cu (40uM) and Zn (125uM) and this test showed similar results to the 40uM Cu continuous test, i.e., no reduction in inhibition. The specific ATP concentration, i.e., ATP/OD, results for the continuous reactor showed an apparent recovery for both Cu-treated populations, although neither the OD nor glucose data showed any recovery. This may reflect that the individual microorganisms that survived after the addition of heavy metals, kept maintaining the usual ATP

  4. Boiling water neutronic reactor incorporating a process inherent safety design

    DOEpatents

    Forsberg, Charles W.

    1987-01-01

    A boiling-water reactor core is positioned within a prestressed concrete reactor vessel of a size which will hold a supply of coolant water sufficient to submerge and cool the reactor core by boiling for a period of at least one week after shutdown. Separate volumes of hot, clean (non-borated) water for cooling during normal operation and cool highly borated water for emergency cooling and reactor shutdown are separated by an insulated wall during normal reactor operation with contact between the two water volumes being maintained at interfaces near the top and bottom ends of the reactor vessel. Means are provided for balancing the pressure of the two volumes at the lower interface zone during normal operation to prevent entry of the cool borated water into the reactor core region, for detecting the onset of excessive power to coolant flow conditions in the reactor core and for detecting low water levels of reactor coolant. Cool borated water is permitted to flow into the reactor core when low reactor coolant levels or excessive power to coolant flow conditions are encountered.

  5. Advanced ceramic cladding for water reactor fuel

    SciTech Connect

    Feinroth, H.

    2000-07-01

    Under the US Department of Energy's Nuclear Energy Research Initiatives (NERI) program, continuous fiber ceramic composites (CFCCs) are being developed as cladding for water reactor fuel elements. The purpose is to substantially increase the passive safety of water reactors. A development effort was initiated in 1991 to fabricate CFCC-clad tubes using commercially available fibers and a sol-gel process developed by McDermott Technologies. Two small-diameter CFCC tubes were fabricated using pure alumina and alumina-zirconia fibers in an alumina matrix. Densities of {approximately}60% of theoretical were achieved. Higher densities are required to guarantee fission gas containment. This NERI work has just begun, and only preliminary results are presented herein. Should the work prove successful, further development is required to evaluate CFCC cladding and performance, including in-pile tests containing fuel and exploring a marriage of CFCC cladding materials with suitable advanced fuel and core designs. The possibility of much higher temperature core designs, possibly cooled with supercritical water, and achievement of plant efficiencies {ge}50% would be examined.

  6. Thermophysical properties of saturated light and heavy water for advanced neutron source applications

    SciTech Connect

    Crabtree, A.; Siman-Tov, M.

    1993-05-01

    The Advanced Neutron Source is an experimental facility being developed by Oak Ridge National Laboratory. As a new nuclear fission research reactor of unprecedented flux, the Advanced Neutron Source Reactor will provide the most intense steady-state beams of neutrons in the world. The high heat fluxes generated in the reactor [303 MW(t) with an average power density of 4.5 MW/L] will be accommodated by a flow of heavy water through the core at high velocities. In support of this experimental and analytical effort, a reliable, highly accurate, and uniform source of thermodynamic and transport property correlations for saturated light and heavy water were developed. In order to attain high accuracy in the correlations, the range of these correlations was limited to the proposed Advanced Neutron Source Reactor's nominal operating conditions. The temperature and corresponding saturation pressure ranges used for light water were 20--300[degrees]C and 0.0025--8.5 MPa, respectively, while those for heavy water were 50--250[degrees]C and 0.012--3.9 MPa. Deviations between the correlation predictions and data from the various sources did not exceed 1.0%. Light water vapor density was the only exception, with an error of 1.76%. The physical property package consists of analytical correlations, SAS codes, and FORTRAN subroutines incorporating these correlations, as well as an interactive, easy-to-use program entitled QuikProp.

  7. Chemical Characterization of Simulated Boiling Water Reactor Coolant

    DTIC Science & Technology

    1990-05-01

    industry to reduce personnel radiation exposure and down-time associated with the operation, mainte- nance and refueling of Light Water Reactor (LWR...AD-A226 654 t t-FILL UIY C CHEMICAL CHARACTERIZATION OF SIMULATED , .BOILING WATER REACTOR COOLANt by Li . . , . , - VERRDON HOLBROOK MASON f ; B.S...CHARACTERIZATION OF SIMULATED BOILING WATER REACTOR COOLANT by VERRDON HOLBROOK MASON Submitted to the Department of Nuclear Engineering on May 9, 1988 in

  8. State space modeling of reactor core in a pressurized water reactor

    SciTech Connect

    Ashaari, A.; Ahmad, T.; M, Wan Munirah W.; Shamsuddin, Mustaffa; Abdullah, M. Adib

    2014-07-10

    The power control system of a nuclear reactor is the key system that ensures a safe operation for a nuclear power plant. However, a mathematical model of a nuclear power plant is in the form of nonlinear process and time dependent that give very hard to be described. One of the important components of a Pressurized Water Reactor is the Reactor core. The aim of this study is to analyze the performance of power produced from a reactor core using temperature of the moderator as an input. Mathematical representation of the state space model of the reactor core control system is presented and analyzed in this paper. The data and parameters are taken from a real time VVER-type Pressurized Water Reactor and will be verified using Matlab and Simulink. Based on the simulation conducted, the results show that the temperature of the moderator plays an important role in determining the power of reactor core.

  9. Environmentally assisted cracking in light water reactors

    SciTech Connect

    Chopra, O.K.; Chung, H.M.; Gruber, E.E.

    1996-07-01

    This report summarizes work performed by Argonne National Laboratory on fatigue and environmentally assisted cracking (EAC) in light water reactors (LWRs) from April 1995 to December 1995. Topics that have been investigated include fatigue of carbon and low-alloy steel used in reactor piping and pressure vessels, EAC of Alloy 600 and 690, and irradiation-assisted stress corrosion cracking (IASCC) of Type 304 SS. Fatigue tests were conducted on ferritic steels in water that contained various concentrations of dissolved oxygen (DO) to determine whether a slow strain rate applied during different portions of a tensile-loading cycle are equally effective in decreasing fatigue life. Crack-growth-rate tests were conducted on compact-tension specimens from several heats of Alloys 600 and 690 in simulated LWR environments. Effects of fluoride-ion contamination on susceptibility to intergranular cracking of high- and commercial- purity Type 304 SS specimens from control-tensile tests at 288 degrees Centigrade. Microchemical changes in the specimens were studied by Auger electron spectroscopy and scanning electron microscopy to determine whether trace impurity elements may contribute to IASCC of these materials.

  10. (Boiling water reactor (BWR) CORA experiments)

    SciTech Connect

    Ott, L.J.

    1990-10-16

    To participate in the 1990 CORA Workshop at Kernforschungszentrum Karlsruhe (KfK) GmbH, Karlsruhe, FRG, on October 1--4, and to participate in detailed discussions on October 5 with the KfK CORA Boiling Water Reactor (BWR) experiments. The traveler attended the 1990 CORA Workshop at KfK, FRG. Participation included the presentation of a paper on work performed by the Boiling Water Reactor Core Melt Progression Phenomena Program at Oak Ridge National Laboratory (ORNL) on posttest analyses of CORA BWR experiments. The Statement of Work (November 1989) for the BWR Core Melt Progression Phenomena Program provides for pretest and posttest analyses of the BWR CORA experiments performed at KfK. Additionally, it is intended that ORNL personnel participate in the planning process for future CORA BWR experiments. For these purposes, meetings were held with KfK staff to discuss such topics as (1) experimental test schedule, (2) BWR test conduct, (3) perceived BWR experimental needs, and (4) KfK operational staff needs with respect to ORNL support. 19 refs.

  11. NUCLEAR SUPERHEATER FOR BOILING WATER REACTOR

    DOEpatents

    Holl, R.J.; Klecker, R.W.; Graham, C.B.

    1962-05-15

    A description is given of a boiling water reactor having a superheating region integral with the core. The core consists essentially of an annular boiling region surrounding an inner superheating region. Both regions contain fuel elements and are separated by a cylindrical wall, perforations being provided in the lower portion of the cylindrical wall to permit circulation of a common water moderator between the two regions. The superheater region comprises a plurality of tubular fuel assemblies through which the steam emanating from the boiling region passes to the steam outlet. Each superheater fuel assembly has an outer double-walled cylinder, the double walls being concentrically spaced and connected together at their upper ends but open at the bottom to provide for differential thermal expansion of the inner and outer walls. Gas is entrapped in the annulus between the walls which acts as an insulating space between the fissionable material inside and the moderator outside. (AEC)

  12. Zeolite Membrane Reactor for Water Gas Shift Reaction for Hydrogen Production

    SciTech Connect

    Lin, Jerry Y.S.

    2013-01-29

    Gasification of biomass or heavy feedstock to produce hydrogen fuel gas using current technology is costly and energy-intensive. The technology includes water gas shift reaction in two or more reactor stages with inter-cooling to maximize conversion for a given catalyst volume. This project is focused on developing a membrane reactor for efficient conversion of water gas shift reaction to produce a hydrogen stream as a fuel and a carbon dioxide stream suitable for sequestration. The project was focused on synthesizing stable, hydrogen perm-selective MFI zeolite membranes for high temperature hydrogen separation; fabricating tubular MFI zeolite membrane reactor and stable water gas shift catalyst for membrane reactor applications, and identifying experimental conditions for water gas shift reaction in the zeolite membrane reactor that will produce a high purity hydrogen stream. The project has improved understanding of zeolite membrane synthesis, high temperature gas diffusion and separation mechanisms for zeolite membranes, synthesis and properties of sulfur resistant catalysts, fabrication and structure optimization of membrane supports, and fundamentals of coupling reaction with separation in zeolite membrane reactor for water gas shift reaction. Through the fundamental study, the research teams have developed MFI zeolite membranes with good perm-selectivity for hydrogen over carbon dioxide, carbon monoxide and water vapor, and high stability for operation in syngas mixture containing 500 part per million hydrogen sulfide at high temperatures around 500°C. The research teams also developed a sulfur resistant catalyst for water gas shift reaction. Modeling and experimental studies on the zeolite membrane reactor for water gas shift reaction have demonstrated the effective use of the zeolite membrane reactor for production of high purity hydrogen stream.

  13. Hydrogen Water Chemistry Technology in Boiling Water Reactors

    SciTech Connect

    Lin, Chien C

    2000-04-15

    Modification of coolant chemistry by feedwater hydrogen addition in boiling water reactors (BWRs), generally called hydrogen water chemistry (HWC), is a viable option to mitigate the intergranular stress corrosion cracking problems for operating BWRs. Some fundamentals in HWC technologies as known today are reviewed. Several full-scale HWC test results are analyzed, and the roles that hydrogen plays in HWC technology are identified and quantitatively evaluated. Some deficiencies in water radiolysis modeling for BWR applications under HWC conditions and the impact of {sup 16}N radiation field increase in the main steam line are also discussed.

  14. [Task 1.] Biodenitrification of low nitrate solar pond waters using sequencing batch reactors. [Task 2.] Solidification/stabilization of high strength and biodenitrified heavy metal sludges with a Portland cement/flyash system

    SciTech Connect

    Figueroa, L.; Cook, N.E.; Siegrist, R.L.; Mosher, J.; Terry, S.; Canonico, S.

    1995-09-22

    Process wastewater and sludges were accumulated on site in solar evaporation ponds during operations at the Department of Energy's Rocky Flats Plant (DOE/RF). Because of the extensive use of nitric acid in the processing of actinide metals, the process wastewater has high concentrations of nitrate. Solar pond waters at DOE/RF contain 300-60,000 mg NO{sub 3}{sup {minus}}/L. Additionally, the pond waters contain varying concentrations of many other aqueous constituents, including heavy metals, alkali salts, carbonates, and low level radioactivity. Solids, both from chemical precipitation and soil material deposition, are also present. Options for ultimate disposal of the pond waters are currently being evaluated and include stabilization and solidification (S/S) by cementation. Removal of nitrates can enhance a wastes amenability to S/S, or can be a unit operation in another treatment scheme. Nitrate removal is also a concern for other sources of pollution at DOE/RF, including contaminated groundwater collected by interceptor trench systems. Finally, nitrate pollution is a problem at many other DOE facilities where actinide metals were processed. The primary objective of this investigation was to optimize biological denitrification of solar pond waters with nitrate concentrations of 300--2,100 mg NO{sub 3}{sup {minus}}/L to below the drinking water standard of 45 mg NO{sub 3}{sup {minus}}/L (10 mg N/L). The effect of pH upon process stability and denitrification rate was determined. In addition, the effect Cr(VI) on denitrification and fate of Cr(VI) in the presence of denitrifying bacteria was evaluated.

  15. Evolution of the core physics concept for the Canadian supercritical water reactor

    SciTech Connect

    Pencer, J.; Colton, A.; Wang, X.; Gaudet, M.; Hamilton, H.; Yetisir, M.

    2013-07-01

    The supercritical water cooled reactor (SCWR) is one of the advanced reactor concepts chosen by the GEN-IV International Forum (GIF) for research and development efforts. Canada's contribution is the Canadian SCWR, a heavy water moderated, pressure tube supercritical light water cooled reactor. Recent developments in the SCWR lattice and core concepts, primarily the introduction of a large central flow tube filled with coolant combined with a two-ring fuel assembly, have enabled significant improvements compared to earlier concepts. These improvements include a reduction in coolant void reactivity (CVR) by more than 10 mk, and an almost 40% increase in fuel exit burnup, which is achieved via balanced power distribution between the fuel pins in the fuel assembly. In this paper the evolution of the physics concept is reviewed, and the present lattice and core physics concepts are presented.

  16. Screening reactor steam/water piping systems for water hammer

    SciTech Connect

    Griffith, P.

    1997-09-01

    A steam/water system possessing a certain combination of thermal, hydraulic and operational states, can, in certain geometries, lead to a steam bubble collapse induced water hammer. These states, operations, and geometries are identified. A procedure that can be used for identifying whether an unbuilt reactor system is prone to water hammer is proposed. For the most common water hammer, steam bubble collapse induced water hammer, six conditions must be met in order for one to occur. These are: (1) the pipe must be almost horizontal; (2) the subcooling must be greater than 20 C; (3) the L/D must be greater than 24; (4) the velocity must be low enough so that the pipe does not run full, i.e., the Froude number must be less than one; (5) there should be void nearby; (6) the pressure must be high enough so that significant damage occurs, that is the pressure should be above 10 atmospheres. Recommendations on how to avoid this kind of water hammer in both the design and the operation of the reactor system are made.

  17. Boiling water reactor licensing basis transient

    SciTech Connect

    Cheng, H. S.; Lu, M. S.; Shier, W. G.; Diamond, D. J.; Levine, M. M.; Odar, F.

    1980-01-01

    An analysis is presented of the licensing basis transient for a boiling water reactor where a turbine trip occurs without steam bypass. The analysis was performed by means of the two-dimensional (R,Z) core dynamics code BNL-TWIGL in conjunction with the system transient code RELAP-3B. Two plant models were used and produced similar results for the analysis of the Peach Bottom turbine trip tests. The models differed in the representation of the steam separator. The analysis of the licensing basis transient produced somewhat different results. The results of sensitivity studies to help explain the differences are presented as well as an analysis of the licensing basis transient with recirculation pump trip. 2 refs., 17 figs., 1 tab.

  18. Commercial Light Water Reactor Tritium Extraction Facility

    SciTech Connect

    McHood, M D

    2000-10-12

    A geotechnical investigation program has been completed for the Commercial Light Water Reactor - Tritium Extraction Facility (CLWR-TEF) at the Savannah River Site (SRS). The program consisted of reviewing previous geotechnical and geologic data and reports, performing subsurface field exploration, field and laboratory testing, and geologic and engineering analyses. The purpose of this investigation was to characterize the subsurface conditions for the CLWR-TEF in terms of subsurface stratigraphy and engineering properties for design and to perform selected engineering analyses. The objectives of the evaluation were to establish site-specific geologic conditions, obtain representative engineering properties of the subsurface and potential fill materials, evaluate the lateral and vertical extent of any soft zones encountered, and perform engineering analyses for slope stability, bearing capacity and settlement, and liquefaction potential. In addition, provide general recommendations for construction and earthwork.

  19. Light water reactor lower head failure analysis

    SciTech Connect

    Rempe, J.L.; Chavez, S.A.; Thinnes, G.L.

    1993-10-01

    This document presents the results from a US Nuclear Regulatory Commission-sponsored research program to investigate the mode and timing of vessel lower head failure. Major objectives of the analysis were to identify plausible failure mechanisms and to develop a method for determining which failure mode would occur first in different light water reactor designs and accident conditions. Failure mechanisms, such as tube ejection, tube rupture, global vessel failure, and localized vessel creep rupture, were studied. Newly developed models and existing models were applied to predict which failure mechanism would occur first in various severe accident scenarios. So that a broader range of conditions could be considered simultaneously, calculations relied heavily on models with closed-form or simplified numerical solution techniques. Finite element techniques-were employed for analytical model verification and examining more detailed phenomena. High-temperature creep and tensile data were obtained for predicting vessel and penetration structural response.

  20. Corrosion problems in light water nuclear reactors

    SciTech Connect

    Berry, W.E.

    1984-06-01

    The corrosion problems encountered during the author's career are reviewed. Attention is given to the development of Zircaloys and attendant factors that affect corrosion; the caustic and chloride stress corrosion cracking (SCC) of austenitic stainless steel steam generator tubing; the qualification of Inconel Alloy 600 for steam generator tubing and the subsequent corrosion problem of secondary side wastage, caustic SCC, pitting, intergranular attack, denting, and primary side SCC; and SCC in weld and furnace sensitized stainless steel piping and internals in boiling water reactor primary coolants. Also mentioned are corrosion of metallic uranium alloy fuels; corrosion of aluminum and niobium candidate fuel element claddings; crevice corrosion and seizing of stainless steel journal-sleeve combinations; SCC of precipitation hardened and martensitic stainless steels; low temperature SCC of welded austenitic stainless steels by chloride, fluoride, and sulfur oxy-anions; and corrosion problems experienced by condensers.

  1. Computer simulation of the NASA water vapor electrolysis reactor

    NASA Technical Reports Server (NTRS)

    Bloom, A. M.

    1974-01-01

    The water vapor electrolysis (WVE) reactor is a spacecraft waste reclamation system for extended-mission manned spacecraft. The WVE reactor's raw material is water, its product oxygen. A computer simulation of the WVE operational processes provided the data required for an optimal design of the WVE unit. The simulation process was implemented with the aid of a FORTRAN IV routine.

  2. Heavy water. January 1972-December 1980 (citations from the International Aerospace abstracts data base). Report for Jan 72-Dec 80. [72 citations

    SciTech Connect

    Not Available

    1980-12-01

    Global information covers studies, applications, testing, and production of heavy water. Discussion of applications intrace studies, deuterium oxide as a moderator in nuclear reactors, and safety regulations for handling heavy water are also included. (Contains 72 citations fully indexed and including a title list.)

  3. Water inventory management in condenser pool of boiling water reactor

    DOEpatents

    Gluntz, Douglas M.

    1996-01-01

    An improved system for managing the water inventory in the condenser pool of a boiling water reactor has means for raising the level of the upper surface of the condenser pool water without adding water to the isolation pool. A tank filled with water is installed in a chamber of the condenser pool. The water-filled tank contains one or more holes or openings at its lowermost periphery and is connected via piping and a passive-type valve (e.g., squib valve) to a high-pressure gas-charged pneumatic tank of appropriate volume. The valve is normally closed, but can be opened at an appropriate time following a loss-of-coolant accident. When the valve opens, high-pressure gas inside the pneumatic tank is released to flow passively through the piping to pressurize the interior of the water-filled tank. In so doing, the initial water contents of the tank are expelled through the openings, causing the water level in the condenser pool to rise. This increases the volume of water available to be boiled off by heat conducted from the passive containment cooling heat exchangers. 4 figs.

  4. Water inventory management in condenser pool of boiling water reactor

    DOEpatents

    Gluntz, D.M.

    1996-03-12

    An improved system for managing the water inventory in the condenser pool of a boiling water reactor has means for raising the level of the upper surface of the condenser pool water without adding water to the isolation pool. A tank filled with water is installed in a chamber of the condenser pool. The water-filled tank contains one or more holes or openings at its lowermost periphery and is connected via piping and a passive-type valve (e.g., squib valve) to a high-pressure gas-charged pneumatic tank of appropriate volume. The valve is normally closed, but can be opened at an appropriate time following a loss-of-coolant accident. When the valve opens, high-pressure gas inside the pneumatic tank is released to flow passively through the piping to pressurize the interior of the water-filled tank. In so doing, the initial water contents of the tank are expelled through the openings, causing the water level in the condenser pool to rise. This increases the volume of water available to be boiled off by heat conducted from the passive containment cooling heat exchangers. 4 figs.

  5. Assessment of released heavy metals from electrical and electronic equipment (EEE) existing in shipwrecks through laboratory-scale simulation reactor.

    PubMed

    Hahladakis, John N; Stylianos, Michailakis; Gidarakos, Evangelos

    2013-04-15

    In a passenger ship, the existence of EEE is obvious. In time, under shipwreck's conditions, all these materials will undergo an accelerated severe corrosion, due to salt water, releasing, consequently, heavy metals and other hazardous substances in the aquatic environment. In this study, a laboratory-scale reactor was manufactured in order to simulate the conditions under which the "Sea Diamond" shipwreck lies (14 bars of pressure and 16°C of temperature) and remotely observe and assess any heavy metal release that would occur, from part of the EEE present in the ship, into the sea. Ten metals were examined and the results showed that zinc, mercury and copper were abundant in the water samples taken from the reactor and in significantly higher concentrations compared to the US EPA CMC (criterion maximum concentration) criterion. Moreover, nickel and lead were found in concentrations higher than the CCC (criterion constant concentration) criterion set by the US EPA for clean seawater. The rest of the elements were measured in concentrations within the permissible limits. It is therefore of environmental benefit to salvage the wreck and recycle all the WEEE found in it.

  6. Transpiring wall supercritical water oxidation test reactor design report

    SciTech Connect

    Haroldsen, B.L.; Ariizumi, D.Y.; Mills, B.E.; Brown, B.G.; Rousar, D.C.

    1996-02-01

    Sandia National Laboratories is working with GenCorp, Aerojet and Foster Wheeler Development Corporation to develop a transpiring wall supercritical water oxidation reactor. The transpiring wall reactor promises to mitigate problems of salt deposition and corrosion by forming a protective boundary layer of pure supercritical water. A laboratory scale test reactor has been assembled to demonstrate the concept. A 1/4 scale transpiring wall reactor was designed and fabricated by Aerojet using their platelet technology. Sandia`s Engineering Evaluation Reactor serves as a test bed to supply, pressurize and heat the waste; collect, measure and analyze the effluent; and control operation of the system. This report describes the design, test capabilities, and operation of this versatile and unique test system with the transpiring wall reactor.

  7. Evaluation of Heavy Metal Removal from Wastewater in a Modified Packed Bed Biofilm Reactor

    PubMed Central

    Azizi, Shohreh; Kamika, Ilunga; Tekere, Memory

    2016-01-01

    For the effective application of a modified packed bed biofilm reactor (PBBR) in wastewater industrial practice, it is essential to distinguish the tolerance of the system for heavy metals removal. The industrial contamination of wastewater from various sources (e.g. Zn, Cu, Cd and Ni) was studied to assess the impacts on a PBBR. This biological system was examined by evaluating the tolerance of different strengths of composite heavy metals at the optimum hydraulic retention time (HRT) of 2 hours. The heavy metal content of the wastewater outlet stream was then compared to the source material. Different biomass concentrations in the reactor were assessed. The results show that the system can efficiently treat 20 (mg/l) concentrations of combined heavy metals at an optimum HRT condition (2 hours), while above this strength there should be a substantially negative impact on treatment efficiency. Average organic reduction, in terms of the chemical oxygen demand (COD) of the system, is reduced above the tolerance limits for heavy metals as mentioned above. The PBBR biological system, in the presence of high surface area carrier media and a high microbial population to the tune of 10 000 (mg/l), is capable of removing the industrial contamination in wastewater. PMID:27186636

  8. Evaluation of Heavy Metal Removal from Wastewater in a Modified Packed Bed Biofilm Reactor.

    PubMed

    Azizi, Shohreh; Kamika, Ilunga; Tekere, Memory

    2016-01-01

    For the effective application of a modified packed bed biofilm reactor (PBBR) in wastewater industrial practice, it is essential to distinguish the tolerance of the system for heavy metals removal. The industrial contamination of wastewater from various sources (e.g. Zn, Cu, Cd and Ni) was studied to assess the impacts on a PBBR. This biological system was examined by evaluating the tolerance of different strengths of composite heavy metals at the optimum hydraulic retention time (HRT) of 2 hours. The heavy metal content of the wastewater outlet stream was then compared to the source material. Different biomass concentrations in the reactor were assessed. The results show that the system can efficiently treat 20 (mg/l) concentrations of combined heavy metals at an optimum HRT condition (2 hours), while above this strength there should be a substantially negative impact on treatment efficiency. Average organic reduction, in terms of the chemical oxygen demand (COD) of the system, is reduced above the tolerance limits for heavy metals as mentioned above. The PBBR biological system, in the presence of high surface area carrier media and a high microbial population to the tune of 10 000 (mg/l), is capable of removing the industrial contamination in wastewater.

  9. Material Removes Heavy Metal Ions From Water

    NASA Technical Reports Server (NTRS)

    Philipp, Warren H., Jr.; Street, Kenneth W.; Hill, Carol; Savino, Joseph M.

    1995-01-01

    New high capacity ion-exchange polymer material removes toxic metal cations from contaminated water. Offers several advantages. High sensitivities for such heavy metals as lead, cadmium, and copper and capable of reducing concentrations in aqueous solutions to parts-per-billion range. Removes cations even when calcium present. Material made into variety of forms, such as thin films, coatings, pellets, and fibers. As result, adapted to many applications to purify contaminated water, usually hard wherever found, whether in wastewater-treatment systems, lakes, ponds, industrial plants, or homes. Another important feature that adsorbed metals easily reclaimed by either destructive or nondestructive process. Other tests show ion-exchange polymer made inexpensively; easy to use; strong, flexible, not easily torn; and chemically stable in storage, in aqueous solutions, and in acidic or basic solution.

  10. Material Removes Heavy Metal Ions From Water

    NASA Technical Reports Server (NTRS)

    Philipp, Warren H., Jr.; Street, Kenneth W.; Hill, Carol; Savino, Joseph M.

    1995-01-01

    New high capacity ion-exchange polymer material removes toxic metal cations from contaminated water. Offers several advantages. High sensitivities for such heavy metals as lead, cadmium, and copper and capable of reducing concentrations in aqueous solutions to parts-per-billion range. Removes cations even when calcium present. Material made into variety of forms, such as thin films, coatings, pellets, and fibers. As result, adapted to many applications to purify contaminated water, usually hard wherever found, whether in wastewater-treatment systems, lakes, ponds, industrial plants, or homes. Another important feature that adsorbed metals easily reclaimed by either destructive or nondestructive process. Other tests show ion-exchange polymer made inexpensively; easy to use; strong, flexible, not easily torn; and chemically stable in storage, in aqueous solutions, and in acidic or basic solution.

  11. Choice of a process design for simultaneous detritiation and upgrading of heavy water for the Advanced Neutron Source

    SciTech Connect

    Miller, A.I.; Spagnolo, D.A.; DeVore, J.R.

    1995-11-01

    Tritium removal and heavy water upgrading are essential components of the heavy water-moderated reactor that is the heart of the Advanced Neutron Source (ANS) to be built at Oak Ridge National Laboratory. The technologies for these two processes, which are closely related, are reviewed in the context of the ANS requirements. The evolution of the design of the Heavy Water Upgrading and Detritiation Facility (HWUDF) for ANS is outlined, and the final conceptual design is presented. The conceptual design of HWUDF has two main component systems: (a) a front-end combined electrolysis and catalytic exchange (CECE) system and (b) a back-end cryogenic distillation (CD) system. The CECE process consists of a countercurrent exchange column for hydrogen-water exchange over a wetproofed catalyst and electrolysis to convert water into hydrogen. It accepts all the tritiated heavy water streams of the reactor and performs an almost total separation into a protium (light hydrogen) stream containing tritium and deuterium at only natural abundance and a deuterium stream containing all the tritium and almost no protium. The tritium-containing deuterium stream is then processed by a CD unit, which removes over 90% of the tritium and concentrates it to >99% tritium for indefinite storage as a metal tritide. Deuterium gas with a small residue of tritium is recombined with oxygen from the electrolytic cells and returned as heavy water to the reactor.

  12. Environmentally assisted cracking in light water reactors.

    SciTech Connect

    Chopra, O. K.; Chung, H. M.; Clark, R. W.; Gruber, E. E.; Shack, W. J.; Soppet, W. K.; Strain, R. V.

    2007-11-06

    This report summarizes work performed by Argonne National Laboratory on fatigue and environmentally assisted cracking (EAC) in light water reactors (LWRs) from January to December 2002. Topics that have been investigated include: (a) environmental effects on fatigue crack initiation in carbon and low-alloy steels and austenitic stainless steels (SSs), (b) irradiation-assisted stress corrosion cracking (IASCC) of austenitic SSs in BWRs, (c) evaluation of causes and mechanisms of irradiation-assisted cracking of austenitic SS in PWRs, and (d) cracking in Ni-alloys and welds. A critical review of the ASME Code fatigue design margins and an assessment of the conservation in the current choice of design margins are presented. The existing fatigue {var_epsilon}-N data have been evaluated to define the effects of key material, loading, and environmental parameters on the fatigue lives of carbon and low-alloy steels and austenitic SSs. Experimental data are presented on the effects of surface roughness on fatigue crack initiation in these materials in air and LWR environments. Crack growth tests were performed in BWR environments on SSs irradiated to 0.9 and 2.0 x 10{sup 21} n x cm{sup -2}. The crack growth rates (CGRs) of the irradiated steels are a factor of {approx}5 higher than the disposition curve proposed in NUREG-0313 for thermally sensitized materials. The CGRs decreased by an order of magnitude in low-dissolved oxygen (DO) environments. Slow-strain-rate tensile (SSRT) tests were conducted in high-purity 289 C water on steels irradiated to {approx}3 dpa. The bulk S content correlated well with the susceptibility to intergranular SCC in 289 C water. The IASCC susceptibility of SSs that contain >0.003 wt. % S increased drastically. bend tests in inert environments at 23 C were conducted on broken pieces of SSRT specimens and on unirradiated specimens of the same materials after hydrogen charging. The results of the tests and a review of other data in the literature

  13. Utilization of Heavy Metal Molten Salts in the ARIES-RS Fusion Reactor

    NASA Astrophysics Data System (ADS)

    Übeyli, Mustafa; Yapıcı, Hüseyin

    2008-09-01

    ARIES-RS is one of the major magnetic fusion energy reactor designs that uses a blanket having vanadium alloy structure cooled by lithium [1, 2]. It is a deuterium-tritium (DT) fusion driven reactor, having a fusion power of 2170 MW [1, 2]. This study presents the neutronic analysis of the ARIES-RS fusion reactor using heavy metal molten salts in which Li2BeF4 as the main constituent was mixed with increased mole fractions of heavy metal salt (ThF4 or UF4) starting by 2 mol.% up to 12 mol.%. Neutron transport calculations were carried out with the help of the SCALE 4.3 system by solving the Boltzmann transport equation with the XSDRNPM code in 238 neutron groups and a S 8- P 3 approximation. According to the numerical results, tritium self-sufficiency was attained for the coolants, Flibe with 2% UF4 or ThF4 and 4% UF4. In addition, higher energy multiplication values were found for the salt with UF4 compared to that with ThF4. Furthermore, significant amount of high quality nuclear fuel was produced to be used in external reactors.

  14. 76 FR 61118 - Meeting of the ACRS Subcommittee on Advanced Boiling Water Reactor; Notice of Meeting

    Federal Register 2010, 2011, 2012, 2013, 2014

    2011-10-03

    ... From the Federal Register Online via the Government Publishing Office NUCLEAR REGULATORY COMMISSION Advisory Committee on Reactor Safeguards (ACRS) Meeting of the ACRS Subcommittee on Advanced Boiling Water Reactor; Notice of Meeting The ACRS Subcommittee on Advanced Boiling Water Reactor...

  15. Water and Regolith Shielding for Surface Reactor Missions

    SciTech Connect

    Poston, David I.; Sadasivan, Pratap; Dixon, David D.; Ade, Brian J.; Leichliter, Katrina J.

    2006-01-20

    This paper investigates potential shielding options for surface power fission reactors. The majority of work is focused on a lunar shield that uses a combination of water in stainless-steel cans and lunar regolith. The major advantage of a water-based shield is that development, testing, and deployment should be relatively inexpensive. This shielding approach is used for three surface reactor concepts: (1) a moderated spectrum, NaK cooled, Hastalloy/UZrH reactor, (2) a fast-spectrum, NaK-cooled, SS/UO2 reactor, and (3) a fast-spectrum, K-heat-pipe-cooled, SS/UO2 reactor. For this study, each of these reactors is coupled to a 25-kWt Stirling power system, designed for 5 year life. The shields are designed to limit the dose both to the Stirling alternators and potential astronauts on the surface. The general configuration used is to bury the reactor, but several other options exist as well. Dose calculations are presented as a function of distance from reactor, depth of buried hole, water boron concentration (if any), and regolith repacked density.

  16. [Heavy metals in water of drinking fountains].

    PubMed

    Segura-Muñoz, Susana I; Trevilato, Tânia M; Takayanagui, Angela M; Hering, Sylvia E; Cupo, Palmira

    2003-03-01

    The purpose of this study was to analyze the levels of lead (Pb), cadmium (Cd), aluminum (Al), zinc (Zn), cooper (Cu) and chromium (Cr) in the water of drinking fountains distributed in the Campus of the University of São Paulo. Ribeirão Preto, Brazil. Thirty random samples were collected in different parts of the Campus that were analyzed by Atomic Absorption Spectrophotometry. According to WHO's Drinking Water Guidelines; lead, cadmium and zinc were found in concentrations higher than those recommended in 40%, 20% and 13% of the samples, respectively. The results were analyzed considering nutritional and toxicological aspects related to the presence of essential and toxic elements for the human being. Reviewing the regulatory limits established in ten countries of America, authors focus in the necessity to find a consensus in the establishment of the higher levels of heavy metals in potable water. The tolerable daily intake, have to be the basis to assure the prevention of diseases associated with a long-term ingestion of these elements through foods.

  17. Process Intensification with Integrated Water-Gas-Shift Membrane Reactor

    SciTech Connect

    2009-11-01

    This factsheet describes a research project whose objective is to develop hydrogen-selective membranes for an innovative gas-separation process based on a water-gas-shift membrane reactor (WGS-MR) for the production of hydrogen.

  18. Advantages of liquid fluoride thorium reactor in comparison with light water reactor

    NASA Astrophysics Data System (ADS)

    Bahri, Che Nor Aniza Che Zainul; Majid, Amran Ab.; Al-Areqi, Wadeeah M.

    2015-04-01

    Liquid Fluoride Thorium Reactor (LFTR) is an innovative design for the thermal breeder reactor that has important potential benefits over the traditional reactor design. LFTR is fluoride based liquid fuel, that use the thorium dissolved in salt mixture of lithium fluoride and beryllium fluoride. Therefore, LFTR technology is fundamentally different from the solid fuel technology currently in use. Although the traditional nuclear reactor technology has been proven, it has perceptual problems with safety and nuclear waste products. The aim of this paper is to discuss the potential advantages of LFTR in three aspects such as safety, fuel efficiency and nuclear waste as an alternative energy generator in the future. Comparisons between LFTR and Light Water Reactor (LWR), on general principles of fuel cycle, resource availability, radiotoxicity and nuclear weapon proliferation shall be elaborated.

  19. Advantages of liquid fluoride thorium reactor in comparison with light water reactor

    SciTech Connect

    Bahri, Che Nor Aniza Che Zainul Majid, Amran Ab.; Al-Areqi, Wadeeah M.

    2015-04-29

    Liquid Fluoride Thorium Reactor (LFTR) is an innovative design for the thermal breeder reactor that has important potential benefits over the traditional reactor design. LFTR is fluoride based liquid fuel, that use the thorium dissolved in salt mixture of lithium fluoride and beryllium fluoride. Therefore, LFTR technology is fundamentally different from the solid fuel technology currently in use. Although the traditional nuclear reactor technology has been proven, it has perceptual problems with safety and nuclear waste products. The aim of this paper is to discuss the potential advantages of LFTR in three aspects such as safety, fuel efficiency and nuclear waste as an alternative energy generator in the future. Comparisons between LFTR and Light Water Reactor (LWR), on general principles of fuel cycle, resource availability, radiotoxicity and nuclear weapon proliferation shall be elaborated.

  20. Modeling of heavy metals removal from municipal landfill leachate using living biomass of water hyacinth.

    PubMed

    el-Gendy, Ahmed S

    2008-01-01

    Water Hyacinth, Eichhornia crassipes, a fast-growing floating aquatic macrophyte; was used for the removal of heavy metals from a municipal landfill leachate. The leachate was spiked with different mixtures of five heavy metals (Cd, Cr, Cu, Pb and Ni), at a range of concentrations to cover the ranges reported in literature. The initial concentrations of the total heavy metals in leachate ranged from 0.06 to 5.5 mequiv L(-1). All experiments were carried out in batch reactors in a greenhouse environment. The water hyacinth plants showed a very promising ability to remove and accumulate these metals from the leachate (24% to 80% removal of total heavy metals). Generally, the reduction in concentration of total heavy metals followed two distinct patterns, a rapid initial decrease followed by a slower decrease. An exponential mathematical model was established to estimate the remaining concentration of total heavy metals in the leachate over time for the rapid initial decrease. Also, a linear relationship was established to estimate the concentration of total heavy metals over time for the slower decrease. In addition, Langmuir and Freundlich isotherms were applied to the observed data and the constants of each isotherm were obtained.

  1. Process for treating effluent from a supercritical water oxidation reactor

    DOEpatents

    Barnes, C.M.; Shapiro, C.

    1997-11-25

    A method for treating a gaseous effluent from a supercritical water oxidation reactor containing entrained solids is provided comprising the steps of expanding the gas/solids effluent from a first to a second lower pressure at a temperature at which no liquid condenses; separating the solids from the gas effluent; neutralizing the effluent to remove any acid gases; condensing the effluent; and retaining the purified effluent to the supercritical water oxidation reactor. 6 figs.

  2. Process for treating effluent from a supercritical water oxidation reactor

    DOEpatents

    Barnes, Charles M.; Shapiro, Carolyn

    1997-01-01

    A method for treating a gaseous effluent from a supercritical water oxidation reactor containing entrained solids is provided comprising the steps of expanding the gas/solids effluent from a first to a second lower pressure at a temperature at which no liquid condenses; separating the solids from the gas effluent; neutralizing the effluent to remove any acid gases; condensing the effluent; and retaining the purified effluent to the supercritical water oxidation reactor.

  3. Stress and Fracture Mechanics Analyses of Boiling Water Reactor and Pressurized Water Reactor Pressure Vessel Nozzles

    SciTech Connect

    Yin, Shengjun; Bass, Bennett Richard; Stevens, Gary; Kirk, Mark

    2011-01-01

    This paper describes stress analysis and fracture mechanics work performed to assess boiling water reactor (BWR) and pressurized water reactor (PWR) nozzles located in the reactor pressure vessel (RPV) adjacent to the core beltline region. Various RPV nozzle geometries were investigated: 1. BWR recirculation outlet nozzle; 2. BWR core spray nozzle3 3. PWR inlet nozzle; ; 4. PWR outlet nozzle; and 5. BWR partial penetration instrument nozzle. The above nozzle designs were selected based on their proximity to the core beltline region, i.e., those nozzle configurations that are located close enough to the core region such that they may receive sufficient fluence prior to end-of-license (EOL) to require evaluation as part of establishing the allowed limits on heatup, cooldown, and hydrotest (leak test) conditions. These nozzles analyzed represent one each of the nozzle types potentially requiring evaluation. The purpose of the analyses performed on these nozzle designs was as follows: To model and understand differences in pressure and thermal stress results using a two-dimensional (2-D) axi-symmetric finite element model (FEM) versus a three-dimensional (3-D) FEM for all nozzle types. In particular, the ovalization (stress concentration) effect of two intersecting cylinders, which is typical of RPV nozzle configurations, was investigated; To verify the accuracy of a selected linear elastic fracture mechanics (LEFM) hand solution for stress intensity factor for a postulated nozzle corner crack for both thermal and pressure loading for all nozzle types; To assess the significance of attached piping loads on the stresses in the nozzle corner region; and To assess the significance of applying pressure on the crack face with respect to the stress intensity factor for a postulated nozzle corner crack.

  4. Advanced Computational Thermal Fluid Physics (CTFP) and Its Assessment for Light Water Reactors and Supercritical Reactors

    SciTech Connect

    D.M. McEligot; K. G. Condie; G. E. McCreery; H. M. McIlroy; R. J. Pink; L.E. Hochreiter; J.D. Jackson; R.H. Pletcher; B.L. Smith; P. Vukoslavcevic; J.M. Wallace; J.Y. Yoo; J.S. Lee; S.T. Ro; S.O. Park

    2005-10-01

    Background: The ultimate goal of the study is the improvement of predictive methods for safety analyses and design of Generation IV reactor systems such as supercritical water reactors (SCWR) for higher efficiency, improved performance and operation, design simplification, enhanced safety and reduced waste and cost. The objective of this Korean / US / laboratory / university collaboration of coupled fundamental computational and experimental studies is to develop the supporting knowledge needed for improved predictive techniques for use in the technology development of Generation IV reactor concepts and their passive safety systems. The present study emphasizes SCWR concepts in the Generation IV program.

  5. Removal of heavy metals and COD by SRB in UAFF reactor

    SciTech Connect

    El Bayoumy, M.; Ali, H.I.; Bewtra, J.K.; Biswas, N.

    1999-06-01

    Sulfate-reducing bacteria, under anaerobic conditions, reduce sulfate, SO{sub 4}{sup {minus}2}, to sulfide, S{sup {minus}2}, which in turn can effectively precipitate heavy metals. In this research project, sulfate-reducing bacteria were grown in an upflow anaerobic fixed-film (UAFF) reactor using optimum growth conditions obtained in previous studies. These reactors were then fed with different heavy metals at increasing loading rates until complete failure occurred as metal removal reached zero and residual sulfide dropped to zero. The metal concentrations were measured as total, dissolved, and free ions both in the influent and in the effluent streams. The results of this research showed that 100% removal efficiencies could be obtained with individual concentrations up to 200 mg/L for Cu, 150 mg/L for Ni and Zn, 75 mg/L for Cr, 50 mg/L for Cd, and 40 mg/L for Pb. Also, the corresponding organic matter removal as total organic carbon was found to be about 50% of the influent total organic carbon. A set of mathematical equations were derived to express the mass balance inside the UAFF reactor, with respect to metal influent concentrations and sulfide production. These equations were corrected by incorporating a correction product, {alpha}{center_dot}{beta}, to represent the toxicity effect of the increasing metal concentrations.

  6. Light Water Reactor Sustainability Accomplishments Report

    SciTech Connect

    McCarthy, Kathryn A.

    2015-02-01

    Welcome to the 2014 Light Water Reactor Sustainability (LWRS) Program Accomplishments Report, covering research and development highlights from 2014. The LWRS Program is a U.S. Department of Energy research and development program to inform and support the long-term operation of our nation’s commercial nuclear power plants. The research uses the unique facilities and capabilities at the Department of Energy national laboratories in collaboration with industry, academia, and international partners. Extending the operating lifetimes of current plants is essential to supporting our nation’s base load energy infrastructure, as well as reaching the Administration’s goal of reducing greenhouse gas emissions to 80% below 1990 levels by the year 2050. The purpose of the LWRS Program is to provide technical results for plant owners to make informed decisions on long-term operation and subsequent license renewal, reducing the uncertainty, and therefore the risk, associated with those decisions. In January 2013, 104 nuclear power plants operated in 31 states. However, since then, five plants have been shut down (several due to economic reasons), with additional shutdowns under consideration. The LWRS Program aims to minimize the number of plants that are shut down, with R&D that supports long-term operation both directly (via data that is needed for subsequent license renewal), as well indirectly (with models and technology that provide economic benefits). The LWRS Program continues to work closely with the Electric Power Research Institute (EPRI) to ensure that the body of information needed to support SLR decisions and actions is available in a timely manner. This report covers selected highlights from the three research pathways in the LWRS Program: Materials Aging and Degradation, Risk-Informed Safety Margin Characterization, and Advanced Instrumentation, Information, and Control Systems Technologies, as well as a look-ahead at planned activities for 2015. If you

  7. Fatigue and environmentally assisted cracking in light water reactors

    SciTech Connect

    Kassner, T.F.; Ruther, W.E.; Chung, H.M.; Hicks, P.D.; Hins, A.G.; Park, J.Y.; Shack, W.J.

    1991-12-01

    Fatigue and environmentally assisted cracking of piping, pressure vessels, and core components in light water reactors (LWRs) are important concerns as extended reactor lifetimes are envisaged. The degradation processes include intergranular stress corrosion cracking (IGSCC) of austenitic stainless steel (SS) piping in boiling water reactors (BWRs), and propagation of fatigue or SCC cracks (which initiate in sensitized SS cladding) into low-alloy ferritic steels in BWR pressure vessels. Similar cracking has also occurred in upper shell-to-transition cone girth welds in pressurized water reactor (PWR) steam generator vessels. Another concern is failure of reactor-core internal components after accumulation of relatively high fluence, which has occurred in both BWRs and PWRs. Research during the past year focused on (1) fatigue and SCC of ferritic steels used in piping and in steam generator and reactor pressure vessels, (2) role of chromate and sulfate in simulated BWR water in SCC of sensitized Type 304 SS, and (3) irradiation-assisted SCC in high- and commercial-purity Type 304 SS specimens from control-blade absorber tubes used in two operating BWRs. Failure after accumulation of relatively high fluence has been attributed to radiation-induced segregation (RIS) of elements such as Si, P, Ni, and Cr. This document provides a summary of research progress in these areas.

  8. REACTOR

    DOEpatents

    Roman, W.G.

    1961-06-27

    A pressurized water reactor in which automatic control is achieved by varying the average density of the liquid moderator-cooiant is patented. Density is controlled by the temperature and power level of the reactor ftself. This control can be effected by the use of either plate, pellet, or tubular fuel elements. The fuel elements are disposed between upper and lower coolant plenum chambers and are designed to permit unrestricted coolant flow. The control chamber has an inlet opening communicating with the lower coolant plenum chamber and a restricted vapor vent communicating with the upper coolant plenum chamber. Thus, a variation in temperature of the fuel elements will cause a variation in the average moderator density in the chamber which directly affects the power level of the reactor.

  9. Issues affecting advanced passive light-water reactor safety analysis

    SciTech Connect

    Beelman, R.J.; Fletcher, C.D.; Modro, S.M.

    1992-08-01

    Next generation commercial reactor designs emphasize enhanced safety through improved safety system reliability and performance by means of system simplification and reliance on immutable natural forces for system operation. Simulating the performance of these safety systems will be central to analytical safety evaluation of advanced passive reactor designs. Yet the characteristically small driving forces of these safety systems pose challenging computational problems to current thermal-hydraulic systems analysis codes. Additionally, the safety systems generally interact closely with one another, requiring accurate, integrated simulation of the nuclear steam supply system, engineered safeguards and containment. Furthermore, numerical safety analysis of these advanced passive reactor designs wig necessitate simulation of long-duration, slowly-developing transients compared with current reactor designs. The composite effects of small computational inaccuracies on induced system interactions and perturbations over long periods may well lead to predicted results which are significantly different than would otherwise be expected or might actually occur. Comparisons between the engineered safety features of competing US advanced light water reactor designs and analogous present day reactor designs are examined relative to the adequacy of existing thermal-hydraulic safety codes in predicting the mechanisms of passive safety. Areas where existing codes might require modification, extension or assessment relative to passive safety designs are identified. Conclusions concerning the applicability of these codes to advanced passive light water reactor safety analysis are presented.

  10. Issues affecting advanced passive light-water reactor safety analysis

    SciTech Connect

    Beelman, R.J.; Fletcher, C.D.; Modro, S.M.

    1992-01-01

    Next generation commercial reactor designs emphasize enhanced safety through improved safety system reliability and performance by means of system simplification and reliance on immutable natural forces for system operation. Simulating the performance of these safety systems will be central to analytical safety evaluation of advanced passive reactor designs. Yet the characteristically small driving forces of these safety systems pose challenging computational problems to current thermal-hydraulic systems analysis codes. Additionally, the safety systems generally interact closely with one another, requiring accurate, integrated simulation of the nuclear steam supply system, engineered safeguards and containment. Furthermore, numerical safety analysis of these advanced passive reactor designs wig necessitate simulation of long-duration, slowly-developing transients compared with current reactor designs. The composite effects of small computational inaccuracies on induced system interactions and perturbations over long periods may well lead to predicted results which are significantly different than would otherwise be expected or might actually occur. Comparisons between the engineered safety features of competing US advanced light water reactor designs and analogous present day reactor designs are examined relative to the adequacy of existing thermal-hydraulic safety codes in predicting the mechanisms of passive safety. Areas where existing codes might require modification, extension or assessment relative to passive safety designs are identified. Conclusions concerning the applicability of these codes to advanced passive light water reactor safety analysis are presented.

  11. An optical dosimeter for monitoring heavy metal ions in water

    NASA Astrophysics Data System (ADS)

    Mignani, Anna G.; Regan, Fiona; Leamy, D.; Mencaglia, A. A.; Ciaccheri, L.

    2005-05-01

    This work presents an optochemical dosimeter for determining and discriminating nickel, copper, and cobalt ions in water that can be used as an early warning system for water pollution. An inexpensive fiber optic spectrophotometer monitors the sensor's spectral behavior under exposure to water solutions of heavy metal ions in the 1-10 mg/l concentration range. The Principal Component Analysis (PCA) method quantitatively determines the heavy metals and discriminates their type and combination.

  12. A Drinking Water Sensor for Lead and Other Heavy Metals.

    PubMed

    Lin, Wen-Chi; Li, Zhongrui; Burns, Mark A

    2017-09-05

    Leakage of lead and other heavy metals into drinking water is a significant health risk and one that is not easily detected. We have developed simple sensors containing only platinum electrodes for the detection of heavy metal contamination in drinking water. The two-electrode sensor can identify the existence of a variety of heavy metals in drinking water, and the four-electrode sensor can distinguish lead from other heavy metals in solution. No false-positive response is generated when the sensors are placed in simulated and actual tap water contaminated by heavy metals. Lead detection on the four-electrode sensor is not affected by the presence of common ions in tap water. Experimental results suggest the sensors can be embedded in water service lines for long-time use until lead or other heavy metals are detected. With its low cost (∼$0.10/sensor) and the possibility of long-term operation, the sensors are ideal for heavy metal detection of drinking water.

  13. Gravity Scaling of a Power Reactor Water Shield

    NASA Technical Reports Server (NTRS)

    Reid, Robert S.; Pearson, J. Boise

    2007-01-01

    A similarity analysis on a water-based reactor shield examined the effect of gravity on free convection between a reactor shield inner and outer vessel boundaries. Two approaches established similarity between operation on the Earth and the Moon: 1) direct scaling of Rayleigh number equating gravity-surface heat flux products, 2) temperature difference between the wall and thermal boundary layer held constant. Nusselt number for natural convection (laminar and turbulent) is assumed of form Nu = CRa(sup n).

  14. LBE water interaction in sub-critical reactors: First experimental and modelling results

    NASA Astrophysics Data System (ADS)

    Ciampichetti, A.; Agostini, P.; Benamati, G.; Bandini, G.; Pellini, D.; Forgione, N.; Oriolo, F.; Ambrosini, W.

    2008-06-01

    This paper concerns the study of the phenomena involved in the interaction between LBE and pressurised water which could occur in some hypothetical accidents in accelerator driven system type reactors. The LIFUS 5 facility was designed and built at ENEA-Brasimone to reproduce this kind of interaction in a wide range of conditions. The first test of the experimental program was carried out injecting water at 70 bar and 235 °C in a reaction vessel containing LBE at 1 bar and 350 °C. A pressurisation up to 80 bar was observed in the test section during the considered transient. The SIMMER III code was used to simulate the performed test. The calculated data agree in a satisfactory way with the experimental results giving confidence in the possibility to use this code for safety analyses of heavy liquid metal cooled reactors.

  15. Advanced Nuclear Technology: Advanced Light Water Reactors Utility Requirements Document Small Modular Reactors Inclusion Summary

    SciTech Connect

    Loflin, Leonard; McRimmon, Beth

    2014-12-18

    This report summarizes a project by EPRI to include requirements for small modular light water reactors (smLWR) into the EPRI Utility Requirements Document (URD) for Advanced Light Water Reactors. The project was jointly funded by EPRI and the U.S. Department of Energy (DOE). The report covers the scope and content of the URD, the process used to revise the URD to include smLWR requirements, a summary of the major changes to the URD to include smLWR, and how to use the URD as revised to achieve value on new plant projects.

  16. Pellet-Cladding Mechanical Interaction Failure Threshold for Reactivity Initiated Accidents for Pressurized Water Reactors and Boiling Water Reactors

    SciTech Connect

    Beyer, Carl E.; Geelhood, Kenneth J.

    2013-06-01

    Pacific Northwest National Laboratory (PNNL) has been requested by the U.S. Nuclear Regulatory Commission to evaluate the reactivity initiated accident (RIA) tests that have recently been performed in the Nuclear Safety Research Reactor (NSRR) and CABRI (French research reactor) on uranium dioxide (UO2) and mixed uranium and plutonium dioxide (MOX) fuels, and to propose pellet-cladding mechanical interaction (PCMI) failure thresholds for RIA events. This report discusses how PNNL developed PCMI failure thresholds for RIA based on least squares (LSQ) regression fits to the RIA test data from cold-worked stress relief annealed (CWSRA) and recrystallized annealed (RXA) cladding alloys under pressurized water reactor (PWR) hot zero power (HZP) conditions and boiling water reactor (BWR) cold zero power (CZP) conditions.

  17. Membrane reactor for water detritiation: a parametric study on operating parameters

    SciTech Connect

    Mascarade, J.; Liger, K.; Troulay, M.; Perrais, C.

    2015-03-15

    This paper presents the results of a parametric study done on a single stage finger-type packed-bed membrane reactor (PBMR) used for heavy water vapor de-deuteration. Parametric studies have been done on 3 operating parameters which are: the membrane temperature, the total feed flow rate and the feed composition through D{sub 2}O content variations. Thanks to mass spectrometer analysis of streams leaving the PBMR, speciation of deuterated species was achieved. Measurement of the amounts of each molecular component allowed the calculation of reaction quotient at the packed-bed outlet. While temperature variation mainly influences permeation efficiency, feed flow rate perturbation reveals dependence of conversion and permeation properties to contact time between catalyst and reacting mixture. The study shows that isotopic exchange reactions occurring on the catalyst particles surface are not thermodynamically balanced. Moreover, the variation of the heavy water content in the feed exhibits competition between permeation and conversion kinetics.

  18. Radiation-initiated removal of heavy metals from water

    SciTech Connect

    Al-Sheikhly, M.; Chaychian, M.; McLaughlin, W.L.

    1995-12-31

    This paper describes the technical viability of high-powered industrial radiation processes as high-volume and low-cost techniques suitable for the removal of the heavy metals and chelated heavy-metal compounds from water that has been contaminated by marine paints and other similar organometallic compounds.

  19. 10 CFR Appendix J to Part 50 - Primary Reactor Containment Leakage Testing for Water-Cooled Power Reactors

    Code of Federal Regulations, 2011 CFR

    2011-01-01

    ... 10 Energy 1 2011-01-01 2011-01-01 false Primary Reactor Containment Leakage Testing for Water-Cooled Power Reactors J Appendix J to Part 50 Energy NUCLEAR REGULATORY COMMISSION DOMESTIC LICENSING OF PRODUCTION AND UTILIZATION FACILITIES Pt. 50, App. J Appendix J to Part 50—Primary Reactor...

  20. Integrity of the reactor coolant boundary of the European pressurized water reactor (EPR)

    SciTech Connect

    Goetsch, D.; Bieniussa, K.; Schulz, H.; Jalouneix, J.

    1997-04-01

    This paper is an abstract of the work performed in the frame of the development of the IPSN/GRS approach in view of the EPR conceptual safety features. EPR is a pressurized water reactor which will be based on the experience gained by utilities and designers in France and in Germany. The reactor coolant boundary of a PWR includes the reactor pressure vessel (RPV), those parts of the steam generators (SGs) which contain primary coolant, the pressurizer (PSR), the reactor coolant pumps (RCPs), the main coolant lines (MCLs) with their branches as well as the other connecting pipes and all branching pipes including the second isolation valves. The present work covering the integrity of the reactor coolant boundary is mainly restricted to the integrity of the main coolant lines (MCLs) and reflects the design requirements for the main components of the reactor coolant boundary. In the following the conceptual aspects, i.e. design, manufacture, construction and operation, will be assessed. A main aspect is the definition of break postulates regarding overall safety implications.

  1. Coupled reactor physics and coolant dynamics of heavy-liquid-metal-coolant systems

    SciTech Connect

    Cahalan, J.E.; Dunn, F.E.; Taiwo, T.A.

    1999-07-01

    Cooling of advanced nuclear designs with heavy liquid metals such as lead or lead-bismuth eutectic offers improved safety characteristics and higher operating efficiencies compared with previously considered liquid-metal coolants such as sodium or NaK. Such applications would, however, also introduce additional safety concerns and design challenges, therefore necessitating a verifiable computational tool for transient design-basis analysis of heavy-liquid-metal-coolant (HLMC) systems. This capability would enable analysts to compare operational and safety characteristics of design alternatives and to evaluate relative performance advantages with a consistent, deterministic measure. No existing computer code provides all the computational capabilities needed for system analysis of HLMC designs. However, the SASSYS-1 computer code, long used for analysis of sodium-cooled designs, has been adapted to provide a scoping reactor kinetics/thermal-hydraulics analysis capability and a framework for future development. Revision of the SASSYS-1 reactor kinetics, thermophysical properties, convective heat transfer, and flow momentum transfer models has permitted preliminary design and safety assessments of HLMC systems. Preliminary results are given here.

  2. Design of virtual SCADA simulation system for pressurized water reactor

    SciTech Connect

    Wijaksono, Umar Abdullah, Ade Gafar; Hakim, Dadang Lukman

    2016-02-08

    The Virtual SCADA system is a software-based Human-Machine Interface that can visualize the process of a plant. This paper described the results of the virtual SCADA system design that aims to recognize the principle of the Nuclear Power Plant type Pressurized Water Reactor. This simulation uses technical data of the Nuclear Power Plant Unit Olkiluoto 3 in Finland. This device was developed using Wonderware Intouch, which is equipped with manual books for each component, animation links, alarm systems, real time and historical trending, and security system. The results showed that in general this device can demonstrate clearly the principles of energy flow and energy conversion processes in Pressurized Water Reactors. This virtual SCADA simulation system can be used as instructional media to recognize the principle of Pressurized Water Reactor.

  3. Design of virtual SCADA simulation system for pressurized water reactor

    NASA Astrophysics Data System (ADS)

    Wijaksono, Umar; Abdullah, Ade Gafar; Hakim, Dadang Lukman

    2016-02-01

    The Virtual SCADA system is a software-based Human-Machine Interface that can visualize the process of a plant. This paper described the results of the virtual SCADA system design that aims to recognize the principle of the Nuclear Power Plant type Pressurized Water Reactor. This simulation uses technical data of the Nuclear Power Plant Unit Olkiluoto 3 in Finland. This device was developed using Wonderware Intouch, which is equipped with manual books for each component, animation links, alarm systems, real time and historical trending, and security system. The results showed that in general this device can demonstrate clearly the principles of energy flow and energy conversion processes in Pressurized Water Reactors. This virtual SCADA simulation system can be used as instructional media to recognize the principle of Pressurized Water Reactor.

  4. Chemical Gradients in Crud on Boiling Water Reactor Fuel Elements

    SciTech Connect

    D. L. Porter; D. E. Janney

    2007-04-01

    Crud (radioactive corrosion products formed inside nuclear reactors is a major problem in commercial power-producing nuclear reactors. Although there are numerous studies of simulated (non-radioactive) crud, characteristics of crud from actual reactors are rarely studied. This study reports scanning electron microscope (SEM) studies of fragments of crud from a commercially operating boiling water reactor. Chemical analyses in the SEM indicated that the crud closest to the outer surfaces of the fuel pins in some areas had Fe:Zn ratios close to 2:1, which decreased away from the fuel pin in some of the fragments. In combination with transmission electron microsope analyses (published elsewhere), these results suggest that the innermost layer of crud in some areas may consist of franklinite (ZnFe2O4, also called zinc spinel), while outer layers in these areas may be predominantly iron oxides.

  5. VIEW OF SOUTHERNMOST OF TWO HEAVY WATER STORAGE TANKS, LOCATED ...

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

    VIEW OF SOUTHERN-MOST OF TWO HEAVY WATER STORAGE TANKS, LOCATED BEHIND SUPPORT COLUMN, WITH ADJACENT PIPING, LEVEL -27’, LOOKING WEST - Physics Assembly Laboratory, Area A/M, Savannah River Site, Aiken, Aiken County, SC

  6. VIEW OF TWO HEAVY WATER STORAGE TANKS (BEHIND SUPPORT COLUMNS ...

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

    VIEW OF TWO HEAVY WATER STORAGE TANKS (BEHIND SUPPORT COLUMNS AND STEEL BEAMS), SUB-BASEMENT LEVEL -27’, LOOKING SOUTHWEST - Physics Assembly Laboratory, Area A/M, Savannah River Site, Aiken, Aiken County, SC

  7. Heavy ion beam degradation from stripping in near vacuum reactor chambers

    SciTech Connect

    Barletta, W.A.

    1981-07-21

    With the use of a particle simulation code we have investigated the ballistic transport of heavy ion beams through a gas-filled reactor for inertial confinement fusion. The background gas pressure has been taken to be 10/sup -4/ torr - 10/sup -3/ torr of Lithium vapor as is appropriate to the HYLIFE reactor concept. During transport to the pellet, Coulomb collisions of beam particles with the background gas will convert a fraction of the beam to charges states higher than the initial value. Collisons will also produce an associated swarm of knock-on electrons. As the beam approaches the pellet, anharmonic components of the space charges forces will lead to a distortion of the phase space of the beam and a consequent degradation of the focal properties of the beam. This degradation can be described in terms of an increase in the rms emittance of the beam. The degree of emittance growth depends sensitivity upon the initial spatial distribution of particles in the beam. For this study we have modified a single-disk particle simulation code, DESTIN (2), to follow two species of particles, the number of which varies in a prescribed fashion dependent upon reactor temperature as the beam converges toward the pellet.

  8. REFLECTOR CONTROL OF A BOILING-WATER REACTOR

    DOEpatents

    Treshow, M.

    1962-05-22

    A line connecting the reactor with a spent steam condenser contains a valve set to open when the pressure in the reactor exceeds a predetermined value and an orifice on the upstream side of the valve. Another line connects the reflector with this line between the orifice and the valve. An excess steam pressure causes the valve to open and the flow of steam through the line draws water out of the reflector. Provision is also made for adding water to the reflector when the steam pressure drops. (AEC)

  9. MEANS FOR SHIELDING REACTORS

    DOEpatents

    Garrison, W.M.; McClinton, L.T.; Burton, M.

    1959-03-10

    A reactor of the heterageneous, heavy water moderated type is described. The reactor is comprised of a plurality of vertically disposed fuel element tubes extending through a tank of heavy water moderator and adapted to accommodate a flow of coolant water in contact with the fuel elements. A tank containing outgoing coolant water is disposed above the core to function is a radiation shield. Unsaturated liquid hydrocarbon is floated on top of the water in the shield tank to reduce to a minimum the possibility of the occurrence of explosive gaseous mixtures resulting from the neutron bombardment of the water in the shield tank.

  10. Pulsed and Continuous UV LED Reactor for Water Treatment

    EPA Pesticide Factsheets

    Numerical data represented in the figures which are graphs.This dataset is associated with the following publication:Spencer, M., M. Miller, J. Richwine, K. Duckworth, L. Racz, M. Grimaila, M. Magnuson , S. Willison , and R. Phillips. Pulsed and Continuous UV LED Reactor for Water Treatment. Aqua - Journal of Water Supply Research and Technology, International Water Supply Association (London, England). Blackwell Publishing, Malden, MA, USA, 1-75, (2016).

  11. Magnetic process for removing heavy metals from water employing magnetites

    DOEpatents

    Prenger, F. Coyne; Hill, Dallas D.; Padilla, Dennis D.; Wingo, Robert M.; Worl, Laura A.; Johnson, Michael D.

    2003-07-22

    A process for removing heavy metals from water is provided. The process includes the steps of introducing magnetite to a quantity of water containing heavy metal. The magnetite is mixed with the water such that at least a portion of, and preferably the majority of, the heavy metal in the water is bound to the magnetite. Once this occurs the magnetite and absorbed metal is removed from the water by application of a magnetic field. In most applications the process is achieved by flowing the water through a solid magnetized matrix, such as steel wool, such that the magnetite magnetically binds to the solid matrix. The magnetized matrix preferably has remnant magnetism, but may also be subject to an externally applied magnetic field. Once the magnetite and associated heavy metal is bound to the matrix, it can be removed and disposed of, such as by reverse water or air and water flow through the matrix. The magnetite may be formed in-situ by the addition of the necessary quantities of Fe(II) and Fe(III) ions, or pre-formed magnetite may be added, or a combination of seed and in-situ formation may be used. The invention also relates to an apparatus for performing the removal of heavy metals from water using the process outlined above.

  12. Magnetic process for removing heavy metals from water employing magnetites

    DOEpatents

    Prenger, F. Coyne; Hill, Dallas D.

    2006-12-26

    A process for removing heavy metals from water is provided. The process includes the steps of introducing magnetite to a quantity of water containing heavy metal. The magnetite is mixed with the water such that at least a portion of, and preferably the majority of, the heavy metal in the water is bound to the magnetite. Once this occurs the magnetite and absorbed metal is removed from the water by application of a magnetic field. In most applications the process is achieved by flowing the water through a solid magnetized matrix, such as steel wool, such that the magnetite magnetically binds to the solid matrix. The magnetized matrix preferably has remnant magnetism, but may also be subject to an externally applied magnetic field. Once the magnetite and associated heavy metal is bound to the matrix, it can be removed and disposed of, such as by reverse water or air and water flow through the matrix. The magnetite may be formed in-situ by the addition of the necessary quantities of Fe(II) and Fe(III) ions, or pre-formed magnetite may be added, or a combination of seed and in-situ formation may be used. The invention also relates to an apparatus for performing the removal of heavy metals from water using the process outlined above.

  13. Self-Sustaining Thorium Boiling Water Reactors

    SciTech Connect

    Greenspan, Ehud; Gorman, Phillip M.; Bogetic, Sandra; Seifried, Jeffrey E.; Zhang, Guanheng; Varela, Christopher R.; Fratoni, Massimiliano; Vijic, Jasmina J.; Downar, Thomas; Hall, Andrew; Ward, Andrew; Jarrett, Michael; Wysocki, Aaron; Xu, Yunlin; Kazimi, Mujid; Shirvan, Koroush; Mieloszyk, Alexander; Todosow, Michael; Brown, Nicolas; Cheng, Lap

    2015-03-15

    The primary objectives of this project are to: Perform a pre-conceptual design of a core for an alternative to the Hitachi proposed fuel-self- sustaining RBWR-AC, to be referred to as a RBWR-Th. The use of thorium fuel is expected to assure negative void coefficient of reactivity (versus positive of the RBWR-AC) and improve reactor safety; Perform a pre-conceptual design of an alternative core to the Hitachi proposed LWR TRU transmuting RBWR-TB2, to be referred to as the RBWR-TR. In addition to improved safety, use of thorium for the fertile fuel is expected to improve the TRU transmutation effectiveness; Compare the RBWR-Th and RBWR-TR performance against that of the Hitachi RBWR core designs and sodium cooled fast reactor counterparts - the ARR and ABR; and, Perform a viability assessment of the thorium-based RBWR design concepts to be identified along with their associated fuel cycle, a technology gap analysis, and a technology development roadmap. A description of the work performed and of the results obtained is provided in this Overview Report and, in more detail, in the Attachments. The major findings of the study are summarized.

  14. Fuel Summary Report: Shippingport Light Water Breeder Reactor - Rev. 2

    SciTech Connect

    Olson, Gail Lynn; Mc Cardell, Richard Keith; Illum, Douglas Brent

    2002-09-01

    The Shippingport Light Water Breeder Reactor (LWBR) was developed by Bettis Atomic Power Laboratory to demonstrate the potential of a water-cooled, thorium oxide fuel cycle breeder reactor. The LWBR core operated from 1977-82 without major incident. The fuel and fuel components suffered minimal damage during operation, and the reactor testing was deemed successful. Extensive destructive and nondestructive postirradiation examinations confirmed that the fuel was in good condition with minimal amounts of cladding deformities and fuel pellet cracks. Fuel was placed in wet storage upon arrival at the Expended Core Facility, then dried and sent to the Idaho Nuclear Technology and Engineering Center for underground dry storage. It is likely that the fuel remains in good condition at its current underground dry storage location at the Idaho Nuclear Technology and Engineering Center. Reports show no indication of damage to the core associated with shipping, loading, or storage.

  15. Assessment of light water reactor accident management programs and experience

    SciTech Connect

    Hammersley, R.J.

    1992-03-01

    The objective of this report is to provide an assessment of the current light water reactor experience regarding accident management programs and associated technology developments. This assessment for light water reactor (LWR) designs is provided as a resource and reference for the development of accident management capabilities for the production reactors at the Savannah River Site. The specific objectives of this assessment are as follows: 1. Perform a review of the NRC, utility, and industry (NUMARC, EPRI) accident management programs and implementation experience. 2. Provide an assessment of the problems and opportunities in developing an accident management program in conjunction or following the Individual Plant Examination process. 3. Review current NRC, utility, and industry technological developments in the areas of computational tools, severe accident predictive tools, diagnostic aids, and severe accident training and simulation.

  16. Sensitivity Analysis of Reprocessing Cooling Times on Light Water Reactor and Sodium Fast Reactor Fuel Cycles

    SciTech Connect

    R. M. Ferrer; S. Bays; M. Pope

    2008-04-01

    The purpose of this study is to quantify the effects of variations of the Light Water Reactor (LWR) Spent Nuclear Fuel (SNF) and fast reactor reprocessing cooling time on a Sodium Fast Reactor (SFR) assuming a single-tier fuel cycle scenario. The results from this study show the effects of different cooling times on the SFR’s transuranic (TRU) conversion ratio (CR) and transuranic fuel enrichment. Also, the decay heat, gamma heat and neutron emission of the SFR’s fresh fuel charge were evaluated. A 1000 MWth commercial-scale SFR design was selected as the baseline in this study. Both metal and oxide CR=0.50 SFR designs are investigated.

  17. Water hyacinth as indicator of heavy metal pollution the tropics

    SciTech Connect

    Gonzalez, H.; Otero, M. ); Lodenius, M. )

    1989-12-01

    The water hyacinth (Eichhornia crassipes) is a common aquatic plant in many tropical countries. Its ability absorb nutrients and other elements from the water has made it possible to use it for water purification purposes. Eichhornia, especially stems and leaves, have been successfully used as indicators of heavy metal pollution in tropical countries. The uptake of heavy metals in this plant is stronger in the roots than in the floating shoots. Metallothionein-like compounds have been found from roots of this species after cadmium exposure. The purpose of this investigation was to study the possibilities of using roots of water hyacinth as a biological indicator of metal pollution in tropical aquatic ecosystems.

  18. Isotopic signature of atmospheric xenon released from light water reactors.

    PubMed

    Kalinowski, Martin B; Pistner, Christoph

    2006-01-01

    A global monitoring system for atmospheric xenon radioactivity is being established as part of the International Monitoring System to verify compliance with the Comprehensive Nuclear-Test-Ban Treaty (CTBT). The isotopic activity ratios of (135)Xe, (133m)Xe, (133)Xe and (131m)Xe are of interest for distinguishing nuclear explosion sources from civilian releases. Simulations of light water reactor (LWR) fuel burn-up through three operational reactor power cycles are conducted to explore the possible xenon isotopic signature of nuclear reactor releases under different operational conditions. It is studied how ratio changes are related to various parameters including the neutron flux, uranium enrichment and fuel burn-up. Further, the impact of diffusion and mixing on the isotopic activity ratio variability are explored. The simulations are validated with reported reactor emissions. In addition, activity ratios are calculated for xenon isotopes released from nuclear explosions and these are compared to the reactor ratios in order to determine whether the discrimination of explosion releases from reactor effluents is possible based on isotopic activity ratios.

  19. Mechanical design of a light water breeder reactor

    DOEpatents

    Fauth, Jr., William L.; Jones, Daniel S.; Kolsun, George J.; Erbes, John G.; Brennan, John J.; Weissburg, James A.; Sharbaugh, John E.

    1976-01-01

    In a light water reactor system using the thorium-232 -- uranium-233 fuel system in a seed-blanket modular core configuration having the modules arranged in a symmetrical array surrounded by a reflector blanket region, the seed regions are disposed for a longitudinal movement between the fixed or stationary blanket region which surrounds each seed region. Control of the reactor is obtained by moving the inner seed region thus changing the geometry of the reactor, and thereby changing the leakage of neutrons from the relatively small seed region into the blanket region. The mechanical design of the Light Water Breeder Reactor (LWBR) core includes means for axially positioning of movable fuel assemblies to achieve the neutron economy required of a breeder reactor, a structure necessary to adequately support the fuel modules without imposing penalties on the breeding capability, a structure necessary to support fuel rods in a closely packed array and a structure necessary to direct and control the flow of coolant to regions in the core in accordance with the heat transfer requirements.

  20. Gravity Scaling of a Power Reactor Water Shield

    SciTech Connect

    Reid, Robert S.; Pearson, J. Boise

    2008-01-21

    Water based reactor shielding is being considered as an affordable option for potential use on initial lunar surface reactor power systems. Heat dissipation in the shield from nuclear sources must be rejected by an auxillary thermal hydraulic cooling system. The mechanism for transferring heat through the shield is natural convection between the core surface and an array of thermosyphon radiator elements. Natural convection in a 100 kWt lunar surface reactor shield design has been previously evaluated at lower power levels (Pearson, 2006). The current baseline assumes that 5.5 kW are dissipated in the water shield, the preponderance on the core surface, but with some volumetric heating in the naturally circulating water as well. This power is rejected by a radiator located above the shield with a surface temperature of 370 K. A similarity analysis on a water-based reactor shield is presented examining the effect of gravity on free convection between a radiation shield inner vessel and a radiation shield outer vessel boundaries. Two approaches established similarity: 1) direct scaling of Rayleigh number equates gravity-surface heat flux products, 2) temperature difference between the wall and thermal boundary layer held constant on Earth and the Moon. Nussult number for natural convection (laminar and turbulent) is assumed of form Nu = CRa{sup n}. These combined results estimate similarity conditions under Earth and Lunar gravities. The influence of reduced gravity on the performance of thermosyphon heat pipes is also examined.

  1. Gravity Scaling of a Power Reactor Water Shield

    NASA Astrophysics Data System (ADS)

    Reid, Robert S.; Pearson, J. Boise

    2008-01-01

    Water based reactor shielding is being considered as an affordable option for potential use on initial lunar surface reactor power systems. Heat dissipation in the shield from nuclear sources must be rejected by an auxillary thermal hydraulic cooling system. The mechanism for transferring heat through the shield is natural convection between the core surface and an array of thermosyphon radiator elements. Natural convection in a 100 kWt lunar surface reactor shield design has been previously evaluated at lower power levels (Pearson, 2006). The current baseline assumes that 5.5 kW are dissipated in the water shield, the preponderance on the core surface, but with some volumetric heating in the naturally circulating water as well. This power is rejected by a radiator located above the shield with a surface temperature of 370 K. A similarity analysis on a water-based reactor shield is presented examining the effect of gravity on free convection between a radiation shield inner vessel and a radiation shield outer vessel boundaries. Two approaches established similarity: 1) direct scaling of Rayleigh number equates gravity-surface heat flux products, 2) temperature difference between the wall and thermal boundary layer held constant on Earth and the Moon. Nussult number for natural convection (laminar and turbulent) is assumed of form Nu = CRan. These combined results estimate similarity conditions under Earth and Lunar gravities. The influence of reduced gravity on the performance of thermosyphon heat pipes is also examined.

  2. Current status of high conversion pressurized water reactor design studies

    SciTech Connect

    Umeoka, T.; Kono, T.; Toyoda, Y.; Ogino, M.; Iwai, S.; Hishida, H.

    1988-01-01

    Preliminary design studies on high conversion pressurized water reactors (HCPWRs) have been completed, and plant design studies are currently being performed to improve the feasibility of HCPWRs. The present status of the feasibility studies is covered, and the related validation tests to be conducted in the coming years are reviewed.

  3. Modeling the behavior of tritiated water vapor in a research reactor containment building.

    PubMed

    Fukui, M

    1992-02-01

    Mathematical models were developed to predict the changes in tritiated water (HTO) concentrations in water pools and of HTO vapor in the Kyoto University Reactor (KUR) containment building in which approximately 3.4 x 10(2) GBq of HTO vapor had leaked from a heavy water facility for more than 1 y. Models reveal that the mechanism of HTO vapor transfer between air and water is controlled by two key parameters: the kinetic constant for HTO exchange and the evaporation rate constant from water to air. A model was constructed based on laboratory experiments using small glass dishes containing various volumes of HTO and was validated by comparing estimates to actual measurements for HTO concentration in water pools of various depths in the containment building. After the leakage from the heavy water facility had been stopped, the decrease in HTO concentration in the sub-pool could be described by this model with a half-life of 15 wk. A mathematical model was also developed to estimate the average HTO vapor concentration in air, which is strongly dependent on the ventilation system's operation even after the removal of the HTO sources. This is due to the continued release of HTO from the concrete material and is analogous to the dynamics of radon emanation. The amount of HTO that has emerged from the concrete was estimated using a model developed for HTO concentration changes in the containment building air, based on long-term monitoring.

  4. Conceptual design of a pressure tube light water reactor with variable moderator control

    SciTech Connect

    Rachamin, R.; Fridman, E.; Galperin, A.

    2012-07-01

    This paper presents the development of innovative pressure tube light water reactor with variable moderator control. The core layout is derived from a CANDU line of reactors in general, and advanced ACR-1000 design in particular. It should be stressed however, that while some of the ACR-1000 mechanical design features are adopted, the core design basics of the reactor proposed here are completely different. First, the inter fuel channels spacing, surrounded by the calandria tank, contains a low pressure gas instead of heavy water moderator. Second, the fuel channel design features an additional/external tube (designated as moderator tube) connected to a separate moderator management system. The moderator management system is design to vary the moderator tube content from 'dry' (gas) to 'flooded' (light water filled). The dynamic variation of the moderator is a unique and very important feature of the proposed design. The moderator variation allows an implementation of the 'breed and burn' mode of operation. The 'breed and burn' mode of operation is implemented by keeping the moderator tube empty ('dry' filled with gas) during the breed part of the fuel depletion and subsequently introducing the moderator by 'flooding' the moderator tube for the 'burn' part. This paper assesses the conceptual feasibility of the proposed concept from a neutronics point of view. (authors)

  5. Experimental limits on the proton life-time from the neutrino experiments with heavy water

    NASA Astrophysics Data System (ADS)

    Tretyak, V. I.; Zdesenko, Y. G.

    2001-04-01

    Experimental data on the number of neutrons born in the heavy water targets of the large neutrino detectors are used to set the limit on the proton life-time independently on decay mode through the reaction d-->n+?. The best up-to-date limit τp>4×1023 yr with 95% C.L. is derived from the measurements with D2O target (mass 267 kg) installed near the Bugey reactor. This value can be improved by six orders of magnitude with future data accumulated with the SNO detector containing 1000 t of D2O.

  6. Materials Degradation in Light Water Reactors: Life After 60,???

    SciTech Connect

    Busby, Jeremy T; Nanstad, Randy K; Stoller, Roger E; Feng, Zhili; Naus, Dan J

    2008-04-01

    Nuclear reactors present a very harsh environment for components service. Components within a reactor core must tolerate high temperature water, stress, vibration, and an intense neutron field. Degradation of materials in this environment can lead to reduced performance, and in some cases, sudden failure. A recent EPRI-led study interviewed 47 US nuclear utility executives to gauge perspectives on long-term operation of nuclear reactors. Nearly 90% indicated that extensions of reactor lifetimes to beyond 60 years were likely. When polled on the most challenging issues facing further life extension, two-thirds cited plant reliability as the key issue with materials aging and cable/piping as the top concerns for plant reliability. Materials degradation within a nuclear power plant is very complex. There are many different types of materials within the reactor itself: over 25 different metal alloys can be found with can be found within the primary and secondary systems, not to mention the concrete containment vessel, instrumentation and control, and other support facilities. When this diverse set of materials is placed in the complex and harsh environment coupled with load, degradation over an extended life is indeed quite complicated. To address this issue, the USNRC has developed a Progressive Materials Degradation Approach (NUREG/CR-6923). This approach is intended to develop a foundation for appropriate actions to keep materials degradation from adversely impacting component integrity and safety and identify materials and locations where degradation can reasonably be expected in the future. Clearly, materials degradation will impact reactor reliability, availability, and potentially, safe operation. Routine surveillance and component replacement can mitigate these factors, although failures still occur. With reactor life extensions to 60 years or beyond or power uprates, many components must tolerate the reactor environment for even longer times. This may increase

  7. The Consortium for Advanced Simulation of Light Water Reactors

    SciTech Connect

    Ronaldo Szilard; Hongbin Zhang; Doug Kothe; Paul Turinsky

    2011-10-01

    The Consortium for Advanced Simulation of Light Water Reactors (CASL) is a DOE Energy Innovation Hub for modeling and simulation of nuclear reactors. It brings together an exceptionally capable team from national labs, industry and academia that will apply existing modeling and simulation capabilities and develop advanced capabilities to create a usable environment for predictive simulation of light water reactors (LWRs). This environment, designated as the Virtual Environment for Reactor Applications (VERA), will incorporate science-based models, state-of-the-art numerical methods, modern computational science and engineering practices, and uncertainty quantification (UQ) and validation against data from operating pressurized water reactors (PWRs). It will couple state-of-the-art fuel performance, neutronics, thermal-hydraulics (T-H), and structural models with existing tools for systems and safety analysis and will be designed for implementation on both today's leadership-class computers and the advanced architecture platforms now under development by the DOE. CASL focuses on a set of challenge problems such as CRUD induced power shift and localized corrosion, grid-to-rod fretting fuel failures, pellet clad interaction, fuel assembly distortion, etc. that encompass the key phenomena limiting the performance of PWRs. It is expected that much of the capability developed will be applicable to other types of reactors. CASL's mission is to develop and apply modeling and simulation capabilities to address three critical areas of performance for nuclear power plants: (1) reduce capital and operating costs per unit energy by enabling power uprates and plant lifetime extension, (2) reduce nuclear waste volume generated by enabling higher fuel burnup, and (3) enhance nuclear safety by enabling high-fidelity predictive capability for component performance.

  8. Fuel Summary Report: Shippingport Light Water Breeder Reactor

    SciTech Connect

    Illum, D.B.; Olson, G.L.; McCardell, R.K.

    1999-01-01

    The Shippingport Light Water Breeder Reactor (LWBR) was a small water cooled, U-233/Th-232 cycle breeder reactor developed by the Pittsburgh Naval Reactors to improve utilization of the nation's nuclear fuel resources in light water reactors. The LWBR was operated at Shippingport Atomic Power Station (APS), which was a Department of Energy (DOE) (formerly Atomic Energy Commission)-owned reactor plant. Shippingport APS was the first large-scale, central-station nuclear power plant in the United States and the first plant of such size in the world operated solely to produce electric power. The Shippingport LWBR was operated successfully from 1977 to 1982 at the APS. During the five years of operation, the LWBR generated more than 29,000 effective full power hours (EFPH) of energy. After final shutdown, the 39 core modules of the LWBR were shipped to the Expended Core Facility (ECF) at Naval Reactors Facility at the Idaho National Engineering and Environmental Laboratory (INEEL). At ECF, 12 of the 39 modules were dismantled and about 1000 of more than 17,000 rods were removed from the modules of proof-of-breeding and fuel performance testing. Some of the removed rods were kept at ECF, some were sent to Argonne National Laboratory-West (ANL-W) in Idaho and some to ANL-East in Chicago for a variety of physical, chemical and radiological examinations. All rods and rod sections remaining after the experiments were shipped back to ECF, where modules and loose rods were repackaged in liners for dry storage. In a series of shipments, the liners were transported from ECF to Idaho Nuclear Technology Engineering Center (INTEC), formerly the Idaho Chemical Processing Plant (ICPP). The 47 liners containing the fully-rodded and partially-derodded core modules, the loose rods, and the rod scraps, are now stored in underground dry wells at CPP-749.

  9. 76 FR 14437 - Economic Simplified Boiling Water Reactor Standard Design: GE Hitachi Nuclear Energy; Issuance of...

    Federal Register 2010, 2011, 2012, 2013, 2014

    2011-03-16

    ... From the Federal Register Online via the Government Publishing Office ] NUCLEAR REGULATORY COMMISSION Economic Simplified Boiling Water Reactor Standard Design: GE Hitachi Nuclear Energy; Issuance of... GE Hitachi Nuclear Energy (GEH) for the economic simplified boiling water reactor (ESBWR)...

  10. 78 FR 56752 - Interim Staff Guidance Specific Environmental Guidance for Integral Pressurized Water Reactors...

    Federal Register 2010, 2011, 2012, 2013, 2014

    2013-09-13

    ... From the Federal Register Online via the Government Publishing Office NUCLEAR REGULATORY COMMISSION Interim Staff Guidance Specific Environmental Guidance for Integral Pressurized Water Reactors... and operate integral pressurized water reactors (iPWR). This guidance applies to environmental reviews...

  11. Degradation Of 4-Chlorophenol Using Water Falling Film Dbd Reactor

    NASA Astrophysics Data System (ADS)

    Dojcinovic, B.; Roglic, G.; Obradovic, B. M.; Kuraica, M. M.; Puric, J.; Natic, M.; Tosti, T.; Manojlovic, D.

    2010-07-01

    In this paper we present experimental results of degradation of 4-chlorphenol using falling film dielectric barrier discharge (DBD) reactor. Degradation of 100 ml/L water solution of 4-chlorphenol was examined using catalysts in four different set of conditions: DBD, DBD/H2O2, DBD/TiO2 i DBD/Fe2+. Kinetics of the 4-chlorophenol degrada- tion in several successive passes trough the reactor was monitored using HPLC. Changes in concentrations of carbon that originate from products of degradations like acetic, formic and oxalic acids and changes in concentrations of carbon calculated on basis of degraded 4-chlorophenole are presented.

  12. Environmentally assisted cracking in light water reactors

    SciTech Connect

    Shack, W.J.; Kassner, T.F.; Maiya, P.S.; Park, J.Y.; Ruther, W.E.

    1988-10-01

    Research during the past year focused on (1) stress corrosion cracking (SCC) of austentitic stainless steels (SS), (2) fatigue of Type 316NG SS, and (3) SCC of ferritic steels used in reactor piping, pressure vessels, and steam generators. Stress corrosion cracking studies on austentitic SS explored the critical strains required for crack initiation, the effects of crevice conditions on SCC susceptibility, heat-to-heat variations in SCC susceptibility of Type 316NG and modified Type 347 SS, the effect of heat treatment on the susceptibility of Type 347 SS, threshold stress intensity values for crack growth in Type 316NG SS, and the effects of cuprous ion and several organic salts on the SCC of sensitized Type 304 SS. Crevice conditions were observed to strongly promote SCC. Significant heat-to-heat variations were observed in SCC susceptibility of Types 316NG and 347 SS. No correlation was found between SCC behavior and minor variations in chemical composition. A significant effect of heat treatment was observed in Type 347 SS. A heat that was extremely resistant to SCC after heat treatment at 650/degree/C for 24 h was susceptible to transgranular stress corrosion cracking (TGSCC) in the solution-annealed condition. Although there was no sensitization in either condition, the presence or absence of precipitates and differences in precipitate morphology appear to influence the SCC behavior. 20 refs., 20 figs., 11 tabs.

  13. Fatigue and environmentally assisted cracking in light water reactors

    SciTech Connect

    Kassner, T.F.; Ruther, W.E.; Chung, H.M.; Hicks, P.D.; Hins, A.G.; Park, J.Y.; Shack, W.J.

    1992-03-01

    Fatigue and stress corrosion cracking (SCC) for low-alloy steel used in piping and in steam generator and reactor pressure vessels have been investigated. Fatigue data were obtained on medium-sulfur-content A533-Gr B and A106-Gr B steels in high-purity (HP) deoxygenated water, in simulated pressurized water reactor water, and in air. Analytical studies focused on the behavior of carbon steels in boiling water reactor (BWR) environments. Crack-growth rates of composite fracture-mechanics specimens of A533-Gr B/Inconel-182/Inconel-600 (plated with nickel) and homogeneous specimens of A533-Gr B steel were determined under small-amplitude cyclic loading in HP water with {approx}300 pbb dissolved oxygen. Radiation-induced segregation and irradiation-assisted SCC of Type 304 SS after accumulation of relatively high fluence also have been investigated. Microchemical and microstructural changes in HP and commercial-purity Type 304 SS specimens from control-blade absorber tubes used in two operating BWRs were studied by Auger electron spectroscopy and scanning electron microscopy, and slow-strain-rate tensile tests were conducted on tubular specimens in air and in simulated BWR water at 289{degrees}C.

  14. Computational Fluid Dynamics Analysis of Canadian Supercritical Water Reactor (SCWR)

    NASA Astrophysics Data System (ADS)

    Movassat, Mohammad; Bailey, Joanne; Yetisir, Metin

    2015-11-01

    A Computational Fluid Dynamics (CFD) simulation was performed on the proposed design for the Canadian SuperCritical Water Reactor (SCWR). The proposed Canadian SCWR is a 1200 MW(e) supercritical light-water cooled nuclear reactor with pressurized fuel channels. The reactor concept uses an inlet plenum that all fuel channels are attached to and an outlet header nested inside the inlet plenum. The coolant enters the inlet plenum at 350 C and exits the outlet header at 625 C. The operating pressure is approximately 26 MPa. The high pressure and high temperature outlet conditions result in a higher electric conversion efficiency as compared to existing light water reactors. In this work, CFD simulations were performed to model fluid flow and heat transfer in the inlet plenum, outlet header, and various parts of the fuel assembly. The ANSYS Fluent solver was used for simulations. Results showed that mass flow rate distribution in fuel channels varies radially and the inner channels achieve higher outlet temperatures. At the outlet header, zones with rotational flow were formed as the fluid from 336 fuel channels merged. Results also suggested that insulation of the outlet header should be considered to reduce the thermal stresses caused by the large temperature gradients.

  15. Recycle of LWR (Light Water Reactor) actinides to an IFR (Integral Fast Reactor)

    SciTech Connect

    Pierce, R.D.; Ackerman, J.P.; Johnson, G.K.; Mulcahey, T.P.; Poa, D.S.

    1991-01-01

    A large quantity of actinide elements is present in irradiated Light Water Reactor (LWR) fuel that is stored throughout the world. Because of the high fission-to-capture ratio for the transuranium (TRU) elements with the high-energy neutrons in the metal-fueled Integral Fast Reactor (IFR), that reactor can consume these elements effectively. The stored fuel represents a valuable resource for an expanding application of fast power reactors. In addition, removal of the TRU elements from the spent LWR fuel has the potential for increasing the capacity of a high-level waste facility by reducing the heat loads and increasing the margin of safety in meeting licensing requirements. Argonne National Laboratory (ANL) is developing a pyrochemical process, which is compatible with the IFR fuel cycle, for the recovery of TRU elements from LWR fuel. The proposed product is a metallic actinide ingot, which can be introduced into the electrorefining step of the IFR process. The major objective of the LWR fuel recovery process is high TRU element recovery, with decontamination a secondary issue, because fission product removal is accomplished in the IFR process. The extensive pyrochemical processing studies of the 1960s and 1970s provide a basis for the design of possible processes. Two processes were selected for laboratory-scale investigation. One is based on the Salt Transport Process studied at ANL for mixed-oxide fast reactor fuel, and the other is based on the blanket processing studies done for ANL's second Experimental Breeder Reactor (EBR-2). This paper discusses the two processes and is a status report on the experimental studies. 5 refs., 2 figs., 2 tabs.

  16. Reactor

    DOEpatents

    Evans, Robert M.

    1976-10-05

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

  17. Upper internals arrangement for a pressurized water reactor

    DOEpatents

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

    2013-07-09

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

  18. Waste disposal from the light water reactor fuel cycle

    NASA Astrophysics Data System (ADS)

    Costello, J. M.; Hardy, C. J.

    1981-05-01

    Alternaive nuclear fuel cycles for support of light water reactors are described and wastes containing naturally occurring or artificially produced radioactivity reviewed. General principles and objectives in radioactive waste management are outlined, and methods for their practical application to fuel cycle wastes discussed. Management of wastes from upgrading processes of uranium hexafluoride manufacture and uranium manufacture and enrichment, and, to a lesser extent, nuclear power reactor wastes are discussed. Some estimates of radiological dose commitments and health effects from nuclear power and fuel cycle wastes were made for US conditions. These indicate that the major part of the radiological dose arises from uranium mining and milling, operation of nuclear reactors, and spent fuel reprocessing. However, the total dose from the fuel cycle is estimated to be only a small fraction of that from natural background radiation.

  19. Criticality Safety of Low-Enriched Uranium and High-Enriched Uranium Fuel Elements in Heavy Water Lattices

    SciTech Connect

    Pesic, Milan P

    2003-10-15

    The RB reactor was designed as a natural-uranium, heavy water, nonreflected critical assembly in the Vinca Institute of Nuclear Sciences, Belgrade, Yugoslavia, in 1958. From 1962 until 2002, numerous critical experiments were carried out with low-enriched uranium and high-enriched uranium fuel elements of tubular shape, known as the Russian TVR-S fuel assembly type, placed in various heavy water square lattices within the RB cylindrical aluminum tank. Some of these well-documented experiments were selected, described, evaluated, and accepted for inclusion in the 'International Handbook of Evaluated Criticality Safety Benchmark Experiments', contributing to the preservation of a rather small number of heavy water benchmark critical experiments.

  20. Application of the failure assessment diagram to the evaluation of pressure-temperature limits for a pressurized water reactor

    SciTech Connect

    Yoon, K.K.; Bloom, J.M.; Pavinich, W.A.; Slager, H.W.

    1984-06-01

    The failure assessment diagram approach, an elastic-plastic fracture mechanics procedure based on the J-integral concept, was used in the evaluation of pressure-temperature (P-T) limits for the beltline region of the vessel of a pressurized water reactor. The main objective of this paper is to illustrate the application of an alternate fracture mechanics method for the evaluation of pressure-temperature limits, as allowed by Title 10, Code of Federal Regulation Part 50 (10 CFR 50), Appendix G. The evaluation of P-T limits for the beltline region of a pressurized water reactor vessel was based on the following assumptions: ASME Pressure Vessel and Piping Code, Section III, Appendix G reference flaw End-of-life fluence level in the beltline region Longitudinal flaw in the beltline weld J-resistance material toughness curves obtained from the U.S. Nuclear Regulatory Commission's Heavy Section Steel Technology (HSST) program Other material properties obtained from the Babcock and Wilcox Integrated Reactor Vessel Material Surveillance Program The maximum allowable pressure levels were calculated at 33 time points along the given bulk coolant temperature history representing the normal operation of a pressurized water reactor. The results of the calculations showed that adequate margins of safety on operating pressure for the critical weld in the beltline of the pressurized water reactor vessel are assured.

  1. Multi-Applications Small Light Water Reactor - NERI Final Report

    SciTech Connect

    S. Michale Modro; James E. Fisher; Kevan D. Weaver; Jose N. Reyes, Jr.; John T. Groome; Pierre Babka; Thomas M. Carlson

    2003-12-01

    The Multi-Application Small Light Water Reactor (MASLWR) project was conducted under the auspices of the Nuclear Energy Research Initiative (NERI) of the U.S. Department of Energy (DOE). The primary project objectives were to develop the conceptual design for a safe and economic small, natural circulation light water reactor, to address the economic and safety attributes of the concept, and to demonstrate the technical feasibility by testing in an integral test facility. This report presents the results of the project. After an initial exploratory and evolutionary process, as documented in the October 2000 report, the project focused on developing a modular reactor design that consists of a self-contained assembly with a reactor vessel, steam generators, and containment. These modular units would be manufactured at a single centralized facility, transported by rail, road, and/or ship, and installed as a series of self-contained units. This approach also allows for staged construction of an NPP and ''pull and replace'' refueling and maintenance during each five-year refueling cycle.

  2. Gravity Scaling of a Power Reactor Water Shield

    NASA Technical Reports Server (NTRS)

    Reid, Robert S.; Pearson, J. Boise

    2008-01-01

    Water based reactor shielding is being considered as an affordable option for use on initial lunar surface power systems. Heat dissipation in the shield from nuclear sources must be rejected by an auxiliary thermal hydraulic cooling system. The mechanism for transferring heat through the shield is natural convection between the core surface and an array of thermosyphon radiator elements. Natural convection in a 100 kWt lunar surface reactor shield design has been previously evaluated at lower power levels (Pearson, 2007). The current baseline assumes that 5.5 kW are dissipated in the water shield, the preponderance on the core surface, but with some volumetric heating in the naturally circulating water as well. This power is rejected by a radiator located above the shield with a surface temperature of 370 K. A similarity analysis on a water-based reactor shield is presented examining the effect of gravity on free convection between a radiation shield inner vessel and a radiation shield outer vessel boundaries. Two approaches established similarity: 1) direct scaling of Rayleigh number equates gravity-surface heat flux products, 2) temperature difference between the wall and thermal boundary layer held constant on Earth and the Moon. Nussult number for natural convection (laminar and turbulent) is assumed of form Nu = CRa(sup n). These combined results estimate similarity conditions under Earth and Lunar gravities. The influence of reduced gravity on the performance of thermosyphon heat pipes is also examined.

  3. Silicon carbide composite for light water reactor fuel assembly applications

    NASA Astrophysics Data System (ADS)

    Yueh, Ken; Terrani, Kurt A.

    2014-05-01

    The feasibility of using SiCf-SiCm composites in light water reactor (LWR) fuel designs was evaluated. The evaluation was motivated by the desire to improve fuel performance under normal and accident conditions. The Fukushima accident once again highlighted the need for improved fuel materials that can maintain fuel integrity to higher temperatures for longer periods of time. The review identified many benefits as well as issues in using the material. Issues perceived as presenting the biggest challenges to the concept were identified to be flux gradient induced differential volumetric swelling, fragmentation and thermal shock resistance. The oxidation of silicon and its release into the coolant as silica has been identified as an issue because existing plant systems have limited ability for its removal. Detailed evaluation using available literature data and testing as part of this evaluation effort have eliminated most of the major concerns. The evaluation identified Boiling Water Reactor (BWR) channel, BWR fuel water tube, and Pressurized Water Reactor (PWR) guide tube as feasible applications for SiC composite. A program has been initiated to resolve some of the remaining issues and to generate physical property data to support the design of commercial fuel components.

  4. 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.

  5. Analysis of scrams and forced outages at boiling water reactors

    SciTech Connect

    Earle, R. T.; Sullivan, W. P.; Miller, K. R.; Schwegman, W. J.

    1980-07-01

    This report documents the results of a study of scrams and forced outages at General Electric Boiling Water Reactors (BWRs) operating in the United States. This study was conducted for Sandia Laboratories under a Light Water Reactor Safety Program which it manages for the United States Department of Energy. Operating plant data were used to identify the causes of scrams and forced outages. Causes of scrams and forced outages have been summarized as a function of operating plant and plant age and also ranked according to the number of events per year, outage time per year, and outage time per event. From this ranking, identified potential improvement opportunities were evaluated to determine the associated benefits and impact on plant availability.

  6. Wear resistant zirconium base alloy article for water reactors

    SciTech Connect

    Gillett, J.E.; Shockling, L.A.; Sherwood, D.G.

    1988-03-01

    In a water reactor operating environment, the combination having improved fretting wear resistance is described comprising: an elongated tubular water displacer rod; having a low neutron absorption cross section guide support plates distributed along the length of the water displacer rod; the water displacer rod intersecting the guide support plates through apertures in the guide support plates; the water displacer rod having a plurality of spaced apart annular electrospark deposited coatings, each coating facing the wall of a respective aperture, the electrospark deposited coatings comprising Cr/sub 2/C/sub 3/; wherein the water displacer rod has a tube wall composed of a zirconium base alloy; and wherein the guide support plates are composed of a stainless steel alloy.

  7. Seventh international symposium on environmental degradation of materials in nuclear power systems -- Water reactors: Proceedings and symposium discussions. Volume 1

    SciTech Connect

    Airey, G.; Andresen, P.; Brown, J.

    1995-12-31

    The papers in this volume are divided into the following areas: pressurized water reactors -- primary side; pressurized water reactors -- secondary side; boiling water reactors -- austenitic alloys; and boiling water reactors -- austenitic alloys/water chemistry. Separate abstracts were prepared for most of the individual papers in this volume.

  8. Fabrication of light water reactor tritium targets

    SciTech Connect

    Pilger, J.P.

    1991-11-01

    The mission of the Fabrication Development Task of the Tritium Target Development Project is: to produce a documented technology basis, including specifications and procedures for target rod fabrication; to demonstrate that light water tritium targets can be manufactured at a rate consistent with tritium production requirements; and to develop quality control methods to evaluate target rod components and assemblies, and establish correlations between evaluated characteristics and target rod performance. Many of the target rod components: cladding tubes, end caps, plenum springs, etc., have similar counterparts in LWR fuel rods. High production rate manufacture and inspection of these components has been adequately demonstrated by nuclear fuel rod manufacturers. This summary describes the more non-conventional manufacturing processes and inspection techniques developed to fabricate target rod components whose manufacturability at required production rates had not been previously demonstrated.

  9. REACTOR PRESSURE VESSEL ISSUES FOR THE LIGHT-WATER REACTOR SUSTAINABILITY PROGRAM

    SciTech Connect

    Nanstad, Randy K; Odette, George Robert

    2010-01-01

    The Light Water Reactor Sustainability Program Plan is a collaborative program between the U.S. Department of Energy and the private sector directed at extending the life of the present generation of nuclear power plants to enable operation to at least 80 years. The reactor pressure vessel (RPV) is one of the primary components requiring significant research to enable such long-term operation. There are significant issues that need to be addressed to reduce the uncertainties in regulatory application, such as, 1) high neutron fluence/long irradiation times, and flux effects, 2) material variability, 3) high-nickel materials, 4)specimen size effects and the fracture toughness master curve, etc. The first issue is the highest priority to obtain the data and mechanistic understanding to enable accurate, reliable embrittlement predictions at high fluences. This paper discusses the major issues associated with long-time operation of existing RPVs and the LWRSP plans to address those issues.

  10. Utilizing heavy metal-laden water hyacinth biomass in vermicomposting.

    PubMed

    Tereshchenko, Natalya N; Akimova, Elena E; Pisarchuk, Anna D; Yunusova, Tatyana V; Minaeva, Oksana M

    2015-05-01

    We studied the efficiency of water treatment by water hyacinth (Eichhornia crassipes) from heavy metals (Zn, Cd, Pb, Cu), as well as a possibility of using water hyacinth biomass obtained during treatment for vermicomposting by Eisenia fetida and the vermicompost quality in a model experiment. The results showed that the concentration of heavy metals in the trials with water hyacinth decreased within 35 days. We introduced water hyacinth biomass to the organic substrate for vermicomposting, which promoted a significant weight gain of earthworms and growth in their number, as well as a 1.5- to 3-fold increase in coprolite production. In the trial with 40 % of Eichhornia biomass in the mixture, we observed a 26-fold increase in the number and a 16-fold weight gain of big mature individuals with clitellum; an increase in the number of small individuals 40 times and in the number of cocoons 140 times, as compared to the initial substrate. The utilization of water hyacinth biomass containing heavy metals in the mixture led to a 10-fold increase in the number of adult individuals and cocoons, which was higher than in control. We found out that adding 10 % of Eichhornia biomass to the initial mixture affected slightly the number of microorganisms and their species diversity in the vermicompost. Adding Eichhornia biomass with heavy metals reduced the total number of microorganisms and sharply diminished their species diversity. In all trials, adding water hyacinth in the mixture for vermicomposting had a positive impact on wheat biometric parameters in a 14-day laboratory experiment, even in the trial with heavy metals.

  11. Numerical study of the effects of lamp configuration and reactor wall roughness in an open channel water disinfection UV reactor.

    PubMed

    Sultan, Tipu

    2016-07-01

    This article describes the assessment of a numerical procedure used to determine the UV lamp configuration and surface roughness effects on an open channel water disinfection UV reactor. The performance of the open channel water disinfection UV reactor was numerically analyzed on the basis of the performance indictor reduction equivalent dose (RED). The RED values were calculated as a function of the Reynolds number to monitor the performance. The flow through the open channel UV reactor was modelled using a k-ε model with scalable wall function, a discrete ordinate (DO) model for fluence rate calculation, a volume of fluid (VOF) model to locate the unknown free surface, a discrete phase model (DPM) to track the pathogen transport, and a modified law of the wall to incorporate the reactor wall roughness effects. The performance analysis was carried out using commercial CFD software (ANSYS Fluent 15.0). Four case studies were analyzed based on open channel UV reactor type (horizontal and vertical) and lamp configuration (parallel and staggered). The results show that lamp configuration can play an important role in the performance of an open channel water disinfection UV reactor. The effects of the reactor wall roughness were Reynolds number dependent. The proposed methodology is useful for performance optimization of an open channel water disinfection UV reactor. Copyright © 2016 Elsevier Ltd. All rights reserved.

  12. 10 CFR Appendix J to Part 50 - Primary Reactor Containment Leakage Testing for Water-Cooled Power Reactors

    Code of Federal Regulations, 2013 CFR

    2013-01-01

    ... their normal mode, and need not be vented. Systems that are normally filled with water and operating... returning the reactor to an operating mode requiring containment integrity. For primary reactor containment... surfaces of the containment structures and components shall be performed prior to any Type A test...

  13. Advanced Water-Gas Shift Membrane Reactor

    SciTech Connect

    Sean Emerson; Thomas Vanderspurt; Susanne Opalka; Rakesh Radhakrishnan; Rhonda Willigan

    2009-01-07

    The overall objectives for this project were: (1) to identify a suitable PdCu tri-metallic alloy membrane with high stability and commercially relevant hydrogen permeation in the presence of trace amounts of carbon monoxide and sulfur; and (2) to identify and synthesize a water gas shift catalyst with a high operating life that is sulfur and chlorine tolerant at low concentrations of these impurities. This work successfully achieved the first project objective to identify a suitable PdCu tri-metallic alloy membrane composition, Pd{sub 0.47}Cu{sub 0.52}G5{sub 0.01}, that was selected based on atomistic and thermodynamic modeling alone. The second objective was partially successful in that catalysts were identified and evaluated that can withstand sulfur in high concentrations and at high pressures, but a long operating life was not achieved at the end of the project. From the limited durability testing it appears that the best catalyst, Pt-Re/Ce{sub 0.333}Zr{sub 0.333}E4{sub 0.333}O{sub 2}, is unable to maintain a long operating life at space velocities of 200,000 h{sup -1}. The reasons for the low durability do not appear to be related to the high concentrations of H{sub 2}S, but rather due to the high operating pressure and the influence the pressure has on the WGS reaction at this space velocity.

  14. VERA Core Simulator methodology for pressurized water reactor cycle depletion

    DOE PAGES

    Kochunas, Brendan; Collins, Benjamin; Stimpson, Shane; ...

    2017-01-12

    This paper describes the methodology developed and implemented in the Virtual Environment for Reactor Applications Core Simulator (VERA-CS) to perform high-fidelity, pressurized water reactor (PWR), multicycle, core physics calculations. Depletion of the core with pin-resolved power and nuclide detail is a significant advance in the state of the art for reactor analysis, providing the level of detail necessary to address the problems of the U.S. Department of Energy Nuclear Reactor Simulation Hub, the Consortium for Advanced Simulation of Light Water Reactors (CASL). VERA-CS has three main components: the neutronics solver MPACT, the thermal-hydraulic (T-H) solver COBRA-TF (CTF), and the nuclidemore » transmutation solver ORIGEN. This paper focuses on MPACT and provides an overview of the resonance self-shielding methods, macroscopic-cross-section calculation, two-dimensional/one-dimensional (2-D/1-D) transport, nuclide depletion, T-H feedback, and other supporting methods representing a minimal set of the capabilities needed to simulate high-fidelity models of a commercial nuclear reactor. Results are presented from the simulation of a model of the first cycle of Watts Bar Unit 1. The simulation is within 16 parts per million boron (ppmB) reactivity for all state points compared to cycle measurements, with an average reactivity bias of <5 ppmB for the entire cycle. Comparisons to cycle 1 flux map data are also provided, and the average 2-D root-mean-square (rms) error during cycle 1 is 1.07%. To demonstrate the multicycle capability, a state point at beginning of cycle (BOC) 2 was also simulated and compared to plant data. The comparison of the cycle 2 BOC state has a reactivity difference of +3 ppmB from measurement, and the 2-D rms of the comparison in the flux maps is 1.77%. Lastly, these results provide confidence in VERA-CS’s capability to perform high-fidelity calculations for practical PWR reactor problems.« less

  15. Boiling water reactor radiation shielded Control Rod Drive Housing Supports

    SciTech Connect

    Baversten, B.; Linden, M.J.

    1995-03-01

    The Control Rod Drive (CRD) mechanisms are located in the area below the reactor vessel in a Boiling Water Reactor (BWR). Specifically, these CRDs are located between the bottom of the reactor vessel and above an interlocking structure of steel bars and rods, herein identified as CRD Housing Supports. The CRD Housing Supports are designed to limit the travel of a Control Rod and Control Rod Drive in the event that the CRD vessel attachement went to fail, allowing the CRD to be ejected from the vessel. By limiting the travel of the ejected CRD, the supports prevent a nuclear overpower excursion that could occur as a result of the ejected CRD. The Housing Support structure must be disassembled in order to remove CRDs for replacement or maintenance. The disassembly task can require a significant amount of outage time and personnel radiation exposure dependent on the number and location of the CRDs to be changed out. This paper presents a way to minimize personal radiation exposure through the re-design of the Housing Support structure. The following paragraphs also delineate a method of avoiding the awkward, manual, handling of the structure under the reactor vessel during a CRD change out.

  16. Process for treating effluent from a supercritical water oxidation reactor

    SciTech Connect

    Barnes, C.M.; Shapiro, C.

    1995-12-31

    The present invention relates generally to a method for treating and recycling the effluent from a supercritical water oxidation reactor and more specifically to a method for treating and recycling the effluent by expanding the effluent without extensive cooling. Supercritical water oxidation is the oxidation of fuel, generally waste material, in a body of water under conditions above the thermodynamic critical point of water. The current state of the art in supercritical water oxidation plant effluent treatment is to cool the reactor effluent through heat exchangers or direct quench, separate the cooled liquid into a gas/vapor stream and a liquid/solid stream, expand the separated effluent, and perform additional purification on gaseous, liquid, brine and solid effluent. If acid gases are present, corrosion is likely to occur in the coolers. During expansion, part of the condensed water will revaporize. Vaporization can damage the valves due to cavitation and erosion. The present invention expands the effluent stream without condensing the stream. Radionuclides and suspended solids are more efficiently separated in the vapor phase. By preventing condensation, the acids are kept in the much less corrosive gaseous phase thereby limiting the damage to treatment equipment. The present invention also reduces the external energy consumption, by utilizing the expansion step to also cool the effluent.

  17. 77 FR 3009 - Knowledge and Abilities Catalog for Nuclear Power Plant Operators: Advanced Boiling Water Reactors

    Federal Register 2010, 2011, 2012, 2013, 2014

    2012-01-20

    ... COMMISSION Knowledge and Abilities Catalog for Nuclear Power Plant Operators: Advanced Boiling Water Reactors..., ``Knowledge and Abilities Catalog for Nuclear Power Plant Operators: Advanced Boiling Water Reactors.'' DATES... developed using this Catalog along with the Operator Licensing Examination Standards for Power Reactors...

  18. Evaluation of biogas production from seaweed in batch tests and in UASB reactors combined with the removal of heavy metals.

    PubMed

    Nkemka, Valentine Nkongndem; Murto, Marika

    2010-07-01

    Seaweed can be anaerobically digested for the production of energy-rich methane. However, the use of seaweed digestate as a fertilizer may be restricted because of the high heavy metal content especially cadmium. Reducing the concentration of heavy metals in the digestate will enable its use as a fertilizer. In this laboratory-scale study, the potential of seaweed and its leachate in the production of methane were evaluated in batch tests. The effect of removing the heavy metals from seaweed leachate was evaluated in both batch test and treatment in an upflow anaerobic sludge blanket (UASB) reactor. The heavy metals were removed from seaweed leachate using an imminodiacetic acid (IDA) polyacrylamide cryogel carrier. The methane yield obtained in the anaerobic digestion of seaweed was 0.12 N l CH(4)/g VS(added). The same methane yield was obtained when the seaweed leachate was used for methane production. The IDA-cryogel carrier was efficient in removing Cd(2+), Cu(2+), Ni(2+) and Zn(2+) ions from seaweed leachate. The removal of heavy metals in the seaweed leachate led to a decrease in the methane yield. The maximum sustainable organic loading rate (OLR) attained in the UASB reactor was 20.6 g tCOD/l/day corresponding to a hydraulic retention time (HRT) of 12 h and with a total COD removal efficiency of about 81%. Hydrolysis and treatment with IDA cryogel reduced the heavy metals content in the seaweed leachate before methane production. This study also demonstrated the suitability of the treatment of seaweed leachate in a UASB reactor.

  19. DIRECT-CYCLE, BOILING-WATER NUCLEAR REACTOR

    DOEpatents

    Harrer, J.M.; Fromm, L.W. Jr.; Kolba, V.M.

    1962-08-14

    A direct-cycle boiling-water nuclear reactor is described that employs a closed vessel and a plurality of fuel assemblies, each comprising an outer tube closed at its lower end, an inner tube, fuel rods in the space between the tubes and within the inner tube. A body of water lying within the pressure vessel and outside the fuel assemblies is converted to saturated steam, which enters each fuel assembly at the top and is converted to superheated steam in the fuel assembly while it is passing therethrough first downward through the space between the inner and outer tubes of the fuel assembly and then upward through the inner tube. (AEC)

  20. New generation of NPP with boiling water reactor of improved safety

    SciTech Connect

    Adamov, E.O.; Kuklin, A.N.; Mityaev, Yu.I.; Mikhan, V.I.; Tokarev, Yu.I.; Cherkashov, Yu.M.; Sokolov, I.N.; Iljin, Yu.V.; Pakh, E.E.; Abramov, V.I.

    1993-12-31

    The nuclear power plants with boiling water reactors of improved safety are being developed. There is 26 years of operating experience with the plant VK-50 in Dimitrovgrad. The design and operation of the BWR reactors are described.

  1. Hybrid reactors. [Fuel cycle

    SciTech Connect

    Moir, R.W.

    1980-09-09

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

  2. Behavior of stainless steels in pressurized water reactor primary circuits

    NASA Astrophysics Data System (ADS)

    Féron, D.; Herms, E.; Tanguy, B.

    2012-08-01

    Stainless steels are widely used in primary circuits of pressurized water reactors (PWRs). Operating experience with the various grades of stainless steels over several decades of years has generally been excellent. Nevertheless, stress corrosion failures have been reported in few cases. Two main factors contributing to SCC susceptibility enhancement are investigated in this study: cold work and irradiation. Irradiation is involved in the stress corrosion cracking and corrosion of in-core reactor components in PWR environment. Irradiated assisted stress corrosion cracking (IASCC) is a complex and multi-physics phenomenon for which a predictive modeling able to describe initiation and/or propagation is not yet achieved. Experimentally, development of initiation smart tests and of in situ instrumentation, also in nuclear reactors, is an important axis in order to gain a better understanding of IASCC kinetics. A strong susceptibility for SCC of heavily cold worked austenitic stainless steels is evidenced in hydrogenated primary water typical of PWRs. It is shown that for a given cold-working procedure, SCC susceptibility of austenitic stainless steels materials increases with increasing cold-work. Results have shown also strong influences of the cold work on the oxide layer composition and of the maximum stress on the time to fracture.

  3. Simulation of Water Gas Shift Zeolite Membrane Reactor

    NASA Astrophysics Data System (ADS)

    Makertiharta, I. G. B. N.; Rizki, Z.; Zunita, Megawati; Dharmawijaya, P. T.

    2017-07-01

    The search of alternative energy sources keeps growing from time to time. Various alternatives have been introduced to reduce the use of fossil fuel, including hydrogen. Many pathways can be used to produce hydrogen. Among all of those, the Water Gas Shift (WGS) reaction is the most common pathway to produce high purity hydrogen. The WGS technique faces a downstream processing challenge due to the removal hydrogen from the product stream itself since it contains a mixture of hydrogen, carbon dioxide and also the excess reactants. An integrated process using zeolite membrane reactor has been introduced to improve the performance of the process by selectively separate the hydrogen whilst boosting the conversion. Furthermore, the zeolite membrane reactor can be further improved via optimizing the process condition. This paper discusses the simulation of Zeolite Membrane Water Gas Shift Reactor (ZMWGSR) with variation of process condition to achieve an optimum performance. The simulation can be simulated into two consecutive mechanisms, the reaction prior to the permeation of gases through the zeolite membrane. This paper is focused on the optimization of the process parameters (e.g. temperature, initial concentration) and also membrane properties (e.g. pore size) to achieve an optimum product specification (concentration, purity).

  4. Retrofittable Modifications to Pressurized Water Reactors for Improved Resource Utilization

    SciTech Connect

    1980-10-01

    This report summarizes work performed for the U.S. Arms Control and Disarmament Agency under BOA AC9NX707 (Task Order 80-02), as part of the Agency's continuing program on improved fuel utilization in light water reactors. The objective of the study was to investigate improvements in fuel management and design of water reactors (PWRs) that could potentially increase the utilization of natural uranium resources in a once-through fuel cycle (i.e., without using spent fuel reprocessing and recycle). For the present study, potential improvements were limited to retrofittable concepts, i.e., those which could be modifications to the reactor system or balance of plant. The potential improvements considered were not necessarily restricted to those which might be economical under current uranium ore prices or to those which might be acceptable to the nuclear industry at the present time. A six-month fuel cycle, for example, although technically possible, would be neither economical nor accept able to the industry at the present time. Although all potential improvements are not necessarily compatible with each other, the target objective was to seek a composite system of compatible improvements that, if possible, could increase uranium resource utilization by 30% or more. Economic factors, risks involved in the introduction, and potential licensing concerns are also addressed in the report.

  5. Carbon-14 discharge at three light-water reactors.

    PubMed

    Kunz, C

    1985-07-01

    A long-term-sampling evaluation was made of the quantity, discharge pathway, and chemical form of 14C released from 2 pressurized water reactors (PWR) and 1 boiling water reactor (BWR) in the northeastern United States. For the R. E. Ginna PWR the discharge rate of gaseous 14C was 11.6 Ci/GW(e)-yr. Venting of gas decay tanks accounted for 42%, while 35% was discharged through auxiliary building ventilation and 23% through containment venting. The average chemical composition was 10% as 14CO2, 90% as 14CH4 and other hydrocarbon gases. For the Indian Point Unit 3 PWR, the discharge rate was 9.6 Ci/GW(e)-yr, primarily by pressure-relief venting and purging of the containment air. Venting of gas decay tanks accounted for about 7% of the total released. The chemical species were 26% 14CO2, 74% 14CH4 and other hydrocarbon gases. For the J. A. FitzPatrick BWR, the discharge rate was 12.4 Ci/GW(e)-yr. Approximately 97% of the release was via off-gas discharge, which was about 95% 14CO2. For all 3 reactors the quantity of 14C released with liquid and solid wastes was less than 5% of the gaseous release.

  6. Transpiring wall supercritical water oxidation reactor salt deposition studies

    SciTech Connect

    Haroldsen, B.L.; Mills, B.E.; Ariizumi, D.Y.; Brown, B.G.

    1996-09-01

    Sandia National Laboratories has teamed with Foster Wheeler Development Corp. and GenCorp, Aerojet to develop and evaluate a new supercritical water oxidation reactor design using a transpiring wall liner. In the design, pure water is injected through small pores in the liner wall to form a protective boundary layer that inhibits salt deposition and corrosion, effects that interfere with system performance. The concept was tested at Sandia on a laboratory-scale transpiring wall reactor that is a 1/4 scale model of a prototype plant being designed for the Army to destroy colored smoke and dye at Pine Bluff Arsenal in Arkansas. During the tests, a single-phase pressurized solution of sodium sulfate (Na{sub 2}SO{sub 4}) was heated to supercritical conditions, causing the salt to precipitate out as a fine solid. On-line diagnostics and post-test observation allowed us to characterize reactor performance at different flow and temperature conditions. Tests with and without the protective boundary layer demonstrated that wall transpiration provides significant protection against salt deposition. Confirmation tests were run with one of the dyes that will be processed in the Pine Bluff facility. The experimental techniques, results, and conclusions are discussed.

  7. Study of Pu consumption in Advanced Light Water Reactors. Evaluation of GE Advanced Boiling Water Reactor plants

    SciTech Connect

    Not Available

    1993-05-13

    Timely disposal of the weapons plutonium is of paramount importance to permanently safeguarding this material. GE`s 1300 MWe Advanced Boiling Water Reactor (ABWR) has been designed to utilize fill] core loading of mixed uranium-plutonium oxide fuel. Because of its large core size, a single ABWR reactor is capable of disposing 100 metric tons of plutonium within 15 years of project inception in the spiking mode. The same amount of material could be disposed of in 25 years after the start of the project as spent fuel, again using a single reactor, while operating at 75 percent capacity factor. In either case, the design permits reuse of the stored spent fuel assemblies for electrical energy generation for the remaining life of the plant for another 40 years. Up to 40 percent of the initial plutonium can also be completely destroyed using ABWRS, without reprocessing, either by utilizing six ABWRs over 25 years or by expanding the disposition time to 60 years, the design life of the plants and using two ABWRS. More complete destruction would require the development and testing of a plutonium-base fuel with a non-fertile matrix for an ABWR or use of an Advanced Liquid Metal Reactor (ALMR). The ABWR, in addition, is fully capable of meeting the tritium target production goals with already developed target technology.

  8. Transactions of the nineteenth water reactor safety information meeting

    SciTech Connect

    Weiss, A.J.

    1991-10-01

    This report contains summaries of papers on reactor safety research to be presented at the 19th Water Reactor Safety Information Meeting at the Bethesda Marriott Hotel in Bethesda, Maryland, October 28--30, 1991. The summaries briefly describe the programs and results of nuclear safety research sponsored by the Office of Nuclear Regulatory Research, USNRC. Summaries of invited papers concerning nuclear safety issues from US government laboratories, the electric utilities, the Electric Power Research Institute (EPRI), the nuclear industry, and from the governments and industry in Europe and Japan are also included. The summaries have been compiled in one report to provide a basis for meaningful discussion and information exchange during the course of the meeting, and are given in the order of their presentation in each session. The individual summaries have been cataloged separately.

  9. Neurocontrol of Pressurized Water Reactors in Load-Follow Operations

    SciTech Connect

    Lin Chaung; Shen Chihming

    2000-12-15

    The neurocontrol technique was applied to control a pressurized water reactor (PWR) in load-follow operations. Generalized learning or direct inverse control architecture was adopted in which the neural network was trained off-line to learn the inverse model of the PWR. Two neural network controllers were designed: One provided control rod position, which controlled the axial power distribution, and the other provided the change in boron concentration, which adjusted core total power. An additional feedback controller was designed so that power tracking capability was improved. The time duration between control actions was 15 min; thus, the xenon effect is limited and can be neglected. Therefore, the xenon concentration was not considered as a controller input variable, which simplified controller design. Center target strategy and minimum boron strategy were used to operate the reactor, and the simulation results demonstrated the effectiveness and performance of the proposed controller.

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

    DOEpatents

    Lau, Louis K. S.

    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.

  11. Creep studies for zircaloy life prediction in water reactors

    NASA Astrophysics Data System (ADS)

    Murty, K. Linga

    1999-10-01

    Zirconium alloys, commonly used as cladding tubes in water reactors, undergo complex biaxial creep deformation. The anisotropic nature of these metals makes it relatively complex to predict their dimensional changes in-reactor. These alloys exhibit transients in creep mechanisms as stress levels change. The underlying creep mechanisms and creep anisotropy depend on the alloy composition as well as the thermomechanical treatment. The anisotropic biaxial creep of cold-worked and recrystallized Zircaloy-4 in terms of Hill’s generalized stress formulation is described, and the temperature and stress dependencies of the steady-state creep rate are reviewed. Predictive models that incorporate anelastic strain are used for transient and transients in creep.

  12. Non-linear analysis in Light Water Reactor design

    SciTech Connect

    Rashid, Y.R.; Sharabi, M.N.; Nickell, R.E.; Esztergar, E.P.; Jones, J.W.

    1980-03-01

    The results obtained from a scoping study sponsored by the US Department of Energy (DOE) under the Light Water Reactor (LWR) Safety Technology Program at Sandia National Laboratories are presented. Basically, this project calls for the examination of the hypothesis that the use of nonlinear analysis methods in the design of LWR systems and components of interest include such items as: the reactor vessel, vessel internals, nozzles and penetrations, component support structures, and containment structures. Piping systems are excluded because they are being addressed by a separate study. Essentially, the findings were that nonlinear analysis methods are beneficial to LWR design from a technical point of view. However, the costs needed to implement these methods are the roadblock to readily adopting them. In this sense, a cost-benefit type of analysis must be made on the various topics identified by these studies and priorities must be established. This document is the complete report by ANATECH International Corporation.

  13. Biodegradation of high concentration phenol containing heavy metal ions by functional biofilm in bioelectro-reactor.

    PubMed

    Li, Xin-gang; Wang, Tao; Sun, Jin-sheng; Huang, Xin; Kong, Xiao-song

    2006-01-01

    Functional microorganisms to high concentration phenol containing Cr6+ and Pb2+ were cultured and biofilm was formed on polypropylene packings in bioelectro-reactor. It was found that the biodegradation capability of such biofilm to phenol changed with the applied voltage. Under the optimal electric field conditions (voltage of 3.0 V, electric field of strength 17.7 V/m and current density of 1.98 A/m2), biodegradation efficiency of phenol aof concentration of 1200 mg/L increased 33% compared to the instance without applying electric field. However, voltage had inverse effect on biodegradation, as microorganisms were killed under strong electric field. Voltage had little effect on heavy ions elimination. Higher absorption rate of Cr6+ and Pb2+ was observed when changing pH from acidic to neutral. The experiment results indicated that, after treatment, 10 L phenol of 2400 mg/L was biodegraded completely within 55 h and concentrations of Cr6+ and Pb2+ dropped to less than 1 mg/L within 12 h and 6 h, from initial values of 50 mg/L and 30 mg/L, respectively.

  14. Performance Assessment of ISS Water Processor Assembly Reactor

    NASA Technical Reports Server (NTRS)

    Carter, Layne; Tatara, James; Mason, Rich; OConner, Ed; Bedard, John

    2004-01-01

    Due to modifications to the ISS waste water composition, the concentration of volatile organics has significantly increased in the feed to the Water Processor Assembly (WPA). In parallel, the oxygen supply pressure increased, resulting in a higher flow rate of oxygen to the WPA. Preliminary testing at Hamilton Sund strand indicated that the higher oxygen flow rate would increase the WPA capacity for volatile organics. Following an analysis of the expected waste water composition, personnel at NASA MSFC and Hamilton Sundstrand conducted a test of a flight-like reactor to assess its capacity for the higher organic loads. The results of this test indicate the WPA can accommodate the expected organic load in the ISS waste water with margin.

  15. Performance Assessment of ISS Water Processor Assembly Reactor

    NASA Technical Reports Server (NTRS)

    Carter, Layne; Tatara, James; Mason, Rich; OConner, Ed; Bedard, John

    2004-01-01

    Due to modifications to the ISS waste water composition, the concentration of volatile organics has significantly increased in the feed to the Water Processor Assembly (WPA). In parallel, the oxygen supply pressure increased, resulting in a higher flow rate of oxygen to the WPA. Preliminary testing at Hamilton Sund strand indicated that the higher oxygen flow rate would increase the WPA capacity for volatile organics. Following an analysis of the expected waste water composition, personnel at NASA MSFC and Hamilton Sundstrand conducted a test of a flight-like reactor to assess its capacity for the higher organic loads. The results of this test indicate the WPA can accommodate the expected organic load in the ISS waste water with margin.

  16. Materials Inventory Database for the Light Water Reactor Sustainability Program

    SciTech Connect

    Kazi Ahmed; Shannon M. Bragg-Sitton

    2013-08-01

    Scientific research involves the purchasing, processing, characterization, and fabrication of many sample materials. The history of such materials can become complicated over their lifetime – materials might be cut into pieces or moved to various storage locations, for example. A database with built-in functions to track these kinds of processes facilitates well-organized research. The Material Inventory Database Accounting System (MIDAS) is an easy-to-use tracking and reference system for such items. The Light Water Reactor Sustainability Program (LWRS), which seeks to advance the long-term reliability and productivity of existing nuclear reactors in the United States through multiple research pathways, proposed MIDAS as an efficient way to organize and track all items used in its research. The database software ensures traceability of all items used in research using built-in functions which can emulate actions on tracked items – fabrication, processing, splitting, and more – by performing operations on the data. MIDAS can recover and display the complete history of any item as a simple report. To ensure the database functions suitably for the organization of research, it was developed alongside a specific experiment to test accident tolerant nuclear fuel cladding under the LWRS Advanced Light Water Reactor Nuclear Fuels Pathway. MIDAS kept track of materials used in this experiment from receipt at the laboratory through all processes, test conduct and, ultimately, post-test analysis. By the end of this process, the database proved to be right tool for this program. The database software will help LWRS more efficiently conduct research experiments, from simple characterization tests to in-reactor experiments. Furthermore, MIDAS is a universal tool that any other research team could use to organize their material inventory.

  17. Numerical study of the effects of surface roughness on water disinfection UV reactor.

    PubMed

    Sultan, Tipu; Ahmad, Sarfraz; Cho, Jinsoo

    2016-04-01

    UV reactors are an emerging choice as a big barrier against the pathogens present in drinking water. However, the precise role of reactor's wall roughness for cross flow ultraviolet (CF-UV) and axial flow ultraviolet (AF-UV) water disinfection reactors are unknown. In this paper, the influences of reactor's wall roughness were investigated with a view to identify their role on the performance factors namely dose distribution and reduction equivalent dose (RED). Herein, the relative effects of reactor's wall roughness on the performance of CF-UV and AF-UV reactors were also highlighted. This numerical study is a first step towards the comprehensive analysis of the effects of reactor's wall roughness for UV reactor. A numerical analysis was performed using ANSYS Fluent 15 academic version. The reactor's wall roughness has a significant effect on the RED. We found that the increase in RED is Reynolds number dependent (at lower value of turbulent Reynolds number the effects are remarkable). The effects of reactor's roughness were more pronounced for AF-UV reactor. The simulation results suggest that the study of reactor's wall roughness provides valuable insight to fully understand the effects of reactor's wall roughness and its impact on the flow behavior and other features of CF-UV and AF-UV water disinfection reactors.

  18. [Research of input water ratio's impact on the quality of effluent water from hydrolysis reactor].

    PubMed

    Liang, Kang-Qiang; Xiong, Ya; Qi, Mao-Rong; Lin, Xiu-Jun; Zhu, Min; Song, Ying-Hao

    2012-11-01

    Based on high SS/BOD and low C/N ratio of waste water of municipal wastewater treatment plant, the structure of currently existing hydrolysis reactor was reformed to improve the influent quality. In order to strengthen the sludge hydrolysis and improve effluent water quality, two layers water distributors were set up so that the sludge hydrolysis zone was formed between the two layers distribution. For the purpose of the hydrolysis reactor not only plays the role of the primary sedimentation tank but also improves the effluent water biodegradability, input water ratios of the upper and lower water distributor in the experiment were changed to get the best input water ratio to guide the large-scale application of this sort hydrolysis reactor. Results show, four kinds of input water ratio have varying degrees COD and SS removal efficiency, however, input water ratio for 1 : 1 can substantially increase SCOD/COD ratio and VFA concentration of effluent water compared with the other three input water ratios. To improve the effluent biodegradability, input water ratio for 1 : 1 was chosen for the best input water ratio. That was the ratio of flow of upper distributor was 50%, and the ratio of the lower one was 50%, at this case it can reduce the processing burden of COD and SS for follow-up treatment, but also improve the biodegradability of the effluent.

  19. Solubility of carbohydrates in heavy water.

    PubMed

    Cardoso, Marcus V C; Carvalho, Larissa V C; Sabadini, Edvaldo

    2012-05-15

    The solubility of several mono-(glucose and xylose), di-(sucrose and maltose), tri-(raffinose) and cyclic (α-cyclodextrin) saccharides in H(2)O and in D(2)O were measured over a range of temperatures. The solution enthalpies for the different carbohydrates in the two solvents were determined using the vant' Hoff equation and the values in D(2)O are presented here for the first time. Our findings indicate that the replacement of H(2)O by D(2)O remarkably decreases the solubilities of the less soluble carbohydrates, such as maltose, raffinose and α-cyclodextrin. On the other hand, the more soluble saccharides, glucose, xylose, and sucrose, are practically insensitive to the H/D replacement in water.

  20. Drinking water denitrification by a membrane bio-reactor.

    PubMed

    Nuhogl, Alper; Pekdemir, Turgay; Yildiz, Ergun; Keskinler, Bulent; Akay, Galip

    2002-03-01

    Drinking water denitrification performance of a bench scale membrane bio-reactor (MBR) was investigated as function of hydraulic and biological parameters. The reactor was a stirred tank and operated both in batch and continuous mode. The mixed denitrifying culture used in the batch mode tests was derived from the mixed liquor of a wastewater treatment plant in Erzincan province in Turkey. But the culture used in the continuous mode tests was that obtained from the batch mode tests at the end of the denitrification process. The nitrate contaminated water treated was separated from the mixed liquor suspended solids (MLSS) containing active mixed denitrifying culture and other organic substances by a membrane of 0.2 microm average pore diameter. The results indicated that the use of a membrane module eliminated the need for additional post treatment processes for the removal of MLSS from the product water. Concentration of nitrite and that of MLSS in the membrane effluent was below the detectable limits. Optimum carbon to nitrogen (C/N) ratio was found to be 2.2 in batch mode tests. Depending on the process conditions, it was possible to obtain denitrification capacities based on the reactor effluent and membrane effluent up to 0.18kgm(-3)day(-1) and 2.44 kg m(-2) day2(-1) NO(3-)-N, respectively. The variation of the removal capacity with reactor dilution rate and membrane permeate flux was the same for two different degrees of [MLSS]0/[NO3-N]0 (mass) ratios of 25.15 and 49.33. The present MBR was able to produce a drinking water with NO(3-)-N concentration of less than 4 ppm from a water with NO3-N contamination level of 367 ppm equivalent to a NO(3-)-N load of 0.310 kgm(-3) day(-1). The results showed that MBR system used was able to offer NO(3-)-N removals of up to 98.5%. It was found that the membrane limiting permeate flux increased with increasing MLSS concentration.

  1. Water cooled breeder program summary report (LWBR (Light Water Breeder Reactor) development program)

    SciTech Connect

    Not Available

    1987-10-01

    The purpose of the Department of Energy Water Cooled Breeder Program was to demonstrate pratical breeding in a uranium-233/thorium fueled core while producing electrical energy in a commercial water reactor generating station. A demonstration Light Water Breeder Reactor (LWBR) was successfully operated for more than 29,000 effective full power hours in the Shippingport Atomic Power Station. The reactor operated with an availability factor of 76% and had a gross electrical output of 2,128,943,470 kilowatt hours. Following operation, the expended core was examined and no evidence of any fuel element defects was found. Nondestructive assay of 524 fuel rods determined that 1.39 percent more fissile fuel was present at the end of core life than at the beginning, proving that breeding had occurred. This demonstrates the existence of a vast source of electrical energy using plentiful domestic thorium potentially capable of supplying the entire national need for many centuries. To build on the successful design and operation of the Shippingport Breeder Core and to provide the technology to implement this concept, several reactor designs of large breeders and prebreeders were developed for commercial-sized plants of 900--1000 Mw(e) net. This report summarizes the Water Cooled Breeder Program from its inception in 1965 to its completion in 1987. Four hundred thirty-six technical reports are referenced which document the work conducted as part of this program. This work demonstrated that the Light Water Breeder Reactor is a viable alternative as a PWR replacement in the next generation of nuclear reactors. This transition would only require a minimum of change in design and fabrication of the reactor and operation of the plant.

  2. Detection and characterization of flaws in segments of light water reactor pressure vessels

    SciTech Connect

    Cook, K.V.; Cunningham, R.A. Jr.; McClung, R.W.

    1987-01-01

    Studies have been conducted to determine flaw density in segments cut from light water reactor (LWR) pressure vessels as part of the Oak Ridge National Laboratory's Heavy-Section Steel Technology (HSST) Program. Segments from the Hope Creek Unit 2 vessil and the Pilgrim Unit 2 Vessel were purchased from salvage dealers. Hope Creek was a boiling water reactor (BWR) design and Pilgrim was a pressurized water reactor (PWR) design. Neither were ever placed in service. Objectives were to evaluate these LWR segments for flaws with ultrasonic and liquid penetrant techniques. Both objectives were successfully completed. One significant indication was detected in a Hope Creek seam weld by ultrasonic techniques and characterized by further analyses terminating with destructive correlation. This indication (with a through-wall dimension of approx.6 mm (approx.0.24 in.)) was detected in only 3 m (10 ft) of weldment and offers extremely limited data when compared to the extent of welding even in a single pressure vessel. However, the detection and confirmation of the flaw in the arbitrarily selected sections implies the Marshall report estimates (and others) are nonconservative for such small flaws. No significant indications were detected in the Pilgrim material by ultrasonic techniques. Unfortunately, the Pilgrim segments contained relatively little weldment; thus, we limited our ultrasonic examinations to the cladding and subcladding regions. Fluorescent liquid penetrant inspection of the cladding surfaces for both LWR segments detected no significant indications (i.e., for a total of approximately 6.8 m/sup 2/ (72 ft/sup 2/) of cladding surface).

  3. Fuzzy power control algorithm for a pressurized water reactor

    SciTech Connect

    Hah, Y.J. ); Lee, B.W. )

    1994-05-01

    A fuzzy power control algorithm is presented for automatic reactor power control in a pressurized water reactor (PWR). Automatic power shape control is complicated by the use of control rods with a conventional proportional-integral-differential controller because it is highly coupled with reactivity compensation. Thus, manual shape controls are usually employed even for the limited capability needed for load-following operations including frequency control. In an attempt to achieve automatic power shape control without any design modifications to the core, a fuzzy power control algorithm is proposed. For the fuzzy control, the rule base is formulated based on a multiple-input multiple-output system. The minimum operation rule and the center of area method are implemented for the development of the fuzzy algorithm. The fuzzy power control algorithm has been applied to Yonggwang Nuclear Unit 3. The simulation results show that the fuzzy control can be adapted as a practical control strategy for automatic reactor power control of PWRs during the load-following operations.

  4. High Performance Fuel Desing for Next Generation Pressurized Water Reactors

    SciTech Connect

    Mujid S. Kazimi; Pavel Hejzlar

    2006-01-31

    The use of internally and externally cooled annular fule rods for high power density Pressurized Water Reactors is assessed. The assessment included steady state and transient thermal conditions, neutronic and fuel management requirements, mechanical vibration issues, fuel performance issues, fuel fabrication methods and econmic assessment. The investigation was donducted by a team from MIT, Westinghouse, Gamma Engineering, Framatome ANP, and AECL. The analyses led to the conclusion that raising the power density by 50% may be possible with this advanced fuel. Even at the 150% power level, the fuel temperature would be a few hundred degrees lower than the current fuel temperatre. Significant economic and safety advantages can be obtained by using this fuel in new reactors. Switching to this type of fuel for existing reactors would yield safety advantages, but the economic return is dependent on the duration of plant shutdown to accommodate higher power production. The main feasiblity issue for the high power performance appears to be the potential for uneven splitting of heat flux between the inner and outer fuel surfaces due to premature closure of the outer fuel-cladding gap. This could be overcome by using a very narrow gap for the inner fuel surface and/or the spraying of a crushable zirconium oxide film at the fuel pellet outer surface. An alternative fuel manufacturing approach using vobropacking was also investigated but appears to yield lower than desirable fuel density.

  5. Camera Inspection Arm for Boiling Water Reactors - 13330

    SciTech Connect

    Martin, Scott; Rood, Marc

    2013-07-01

    Boiling Water Reactor (BWR) outage maintenance tasks can be time-consuming and hazardous. Reactor facilities are continuously looking for quicker, safer, and more effective methods of performing routine inspection during these outages. In 2011, S.A. Technology (SAT) was approached by Energy Northwest to provide a remote system capable of increasing efficiencies related to Reactor Pressure Vessel (RPV) internal inspection activities. The specific intent of the system discussed was to inspect recirculation jet pumps in a manner that did not require manual tooling, and could be performed independently of other ongoing inspection activities. In 2012, SAT developed a compact, remote, camera inspection arm to create a safer, more efficient outage environment. This arm incorporates a compact and lightweight design along with the innovative use of bi-stable composite tubes to provide a six-degree of freedom inspection tool capable of reducing dose uptake, reducing crew size, and reducing the overall critical path for jet pump inspections. The prototype camera inspection arm unit is scheduled for final testing in early 2013 in preparation for the Columbia Generating Station refueling outage in the spring of 2013. (authors)

  6. High temperature reactors

    NASA Astrophysics Data System (ADS)

    Dulera, I. V.; Sinha, R. K.

    2008-12-01

    With the advent of high temperature reactors, nuclear energy, in addition to producing electricity, has shown enormous potential for the production of alternate transport energy carrier such as hydrogen. High efficiency hydrogen production processes need process heat at temperatures around 1173-1223 K. Bhabha Atomic Research Centre (BARC), is currently developing concepts of high temperature reactors capable of supplying process heat around 1273 K. These reactors would provide energy to facilitate combined production of hydrogen, electricity, and drinking water. Compact high temperature reactor is being developed as a technology demonstrator for associated technologies. Design has been also initiated for a 600 MWth innovative high temperature reactor. High temperature reactor development programme has opened new avenues for research in areas like advanced nuclear fuels, high temperature and corrosion resistant materials and protective coatings, heavy liquid metal coolant technologies, etc. The paper highlights design of these reactors and their material related requirements.

  7. Supercritical Water Reactor Cycle for Medium Power Applications

    SciTech Connect

    BD Middleton; J Buongiorno

    2007-04-25

    Scoping studies for a power conversion system based on a direct-cycle supercritical water reactor have been conducted. The electric power range of interest is 5-30 MWe with a design point of 20 MWe. The overall design objective is to develop a system that has minimized physical size and performs satisfactorily over a broad range of operating conditions. The design constraints are as follows: Net cycle thermal efficiency {ge}20%; Steam turbine outlet quality {ge}90%; and Pumping power {le}2500 kW (at nominal conditions). Three basic cycle configurations were analyzed. Listed in order of increased plant complexity, they are: (1) Simple supercritical Rankine cycle; (2) All-supercritical Brayton cycle; and (3) Supercritical Rankine cycle with feedwater preheating. The sensitivity of these three configurations to various parameters, such as reactor exit temperature, reactor pressure, condenser pressure, etc., was assessed. The Thermoflex software package was used for this task. The results are as follows: (a) The simple supercritical Rankine cycle offers the greatest hardware simplification, but its high reactor temperature rise and reactor outlet temperature may pose serious problems from the viewpoint of thermal stresses, stability and materials in the core. (b) The all-supercritical Brayton cycle is not a contender, due to its poor thermal efficiency. (c) The supercritical Rankine cycle with feedwater preheating affords acceptable thermal efficiency with lower reactor temperature rise and outlet temperature. (d) The use of a moisture separator improves the performance of the supercritical Rankine cycle with feedwater preheating and allows for a further reduction of the reactor outlet temperature, thus it was selected for the next step. Preliminary engineering design of the supercritical Rankine cycle with feedwater preheating and moisture separation was performed. All major components including the turbine, feedwater heater, feedwater pump, condenser, condenser pump

  8. Characterization of 14C in Swedish light water reactors.

    PubMed

    Magnusson, Asa; Aronsson, Per-Olof; Lundgren, Klas; Stenström, Kristina

    2008-08-01

    This paper presents the results of a 4-y investigation of 14C in different waste streams of both boiling water reactors (BWRs) and pressurized water reactors (PWRs). Due to the potential impact of 14C on human health, minimizing waste and releases from the nuclear power industry is of considerable interest. The experimental data and conclusions may be implemented to select appropriate waste management strategies and practices at reactor units and disposal facilities. Organic and inorganic 14C in spent ion exchange resins, process water systems, ejector off-gas and replaced steam generator tubes were analyzed using a recently developed extraction method. Separate analysis of the chemical species is of importance in order to model and predict the fate of 14C within process systems as well as in dose calculations for disposal facilities. By combining the results of this investigation with newly calculated production rates, mass balance assessments were made of the 14C originating from production in the coolant. Of the 14C formed in the coolant of BWRs, 0.6-0.8% was found to be accumulated in the ion exchange resins (core-specific production rate in the coolant of a 2,500 MWth BWR calculated to be 580 GBq GW(e)(-1) y(-1)). The corresponding value for PWRs was 6-10% (production rate in a 2,775 MWth PWR calculated to be 350 GBq GW(e)(-1) y(-1)). The 14C released with liquid discharges was found to be insignificant, constituting less than 0.5% of the production in the coolant. The stack releases, routinely measured at the power plants, were found to correspond to 60-155% of the calculated coolant production, with large variations between the BWR units.

  9. Neutronic Reactor Structure

    DOEpatents

    Vernon, H. C.; Weinberg, A. M.

    1961-05-30

    The neutronic reactor is comprised of a core consisting of natural uranium and heavy water with a K-factor greater than unity. The core is surrounded by a reflector consisting of natural uranium and ordinary water with a Kfactor less than unity. (AEC)

  10. NEUTRONIC REACTOR STRUCTURE

    DOEpatents

    Weinberg, A.M.; Vernon, H.C.

    1961-05-30

    A neutronic reactor is described. It has a core consisting of natural uranium and heavy water and having a K-factor greater than unity which is surrounded by a reflector consisting of natural uranium and ordinary water having a Kfactor less than unity.

  11. An overview on the reactors to study drinking water biofilms.

    PubMed

    Gomes, I B; Simões, M; Simões, L C

    2014-10-01

    The development of biofilms in drinking water distribution systems (DWDS) can cause pipe degradation, changes in the water organoleptic properties but the main problem is related to the public health. Biofilms are the main responsible for the microbial presence in drinking water (DW) and can be reservoirs for pathogens. Therefore, the understanding of the mechanisms underlying biofilm formation and behavior is of utmost importance in order to create effective control strategies. As the study of biofilms in real DWDS is difficult, several devices have been developed. These devices allow biofilm formation under controlled conditions of physical (flow velocity, shear stress, temperature, type of pipe material, etc), chemical (type and amount of nutrients, type of disinfectant and residuals, organic and inorganic particles, ions, etc) and biological (composition of microbial community - type of microorganism and characteristics) parameters, ensuring that the operational conditions are similar as possible to the DWDS conditions in order to achieve results that can be applied to the real scenarios. The devices used in DW biofilm studies can be divided essentially in two groups, those usually applied in situ and the bench top laboratorial reactors. The selection of a device should be obviously in accordance with the aim of the study and its advantages and limitations should be evaluated to obtain reproducible results that can be transposed into the reality of the DWDS. The aim of this review is to provide an overview on the main reactors used in DW biofilm studies, describing their characteristics and applications, taking into account their main advantages and limitations.

  12. Reducing the cobalt inventory in light water reactors

    SciTech Connect

    Ocken, H.

    1985-01-01

    Reducing the cobalt content of materials used in nuclear power plants is one approach to controlling the radiation fields responsible for occupational radiation exposure; corrosion of steam generator tubing is the primary source in pressurized water reactors (PWRs). Wear of the cobalt-base alloys used to hardface valves (especially feedwater regulator valves) and as pins and rollers in control blades are the primary boiling water reactor (BWR) sources. Routine valve maintenance can also be a significant source of cobalt. Wear, mechanical property, and corrosion measurements led to the selection of Nitronic-60/CFA and PH 13-8 Mo/Inconel X-750 as low-cobalt alloys for use as pin/roller combinations. These alloys are currently being tested in two commercial BWRs. Measurements show that Type 440C stainless steel wears less than the cobalt-base alloys in BWR feedwater regulator valves. Sliding wear tests performed at room temperature in simulated PWR water showed that Colmonoy 74 and 84, Deloro 40, and Vertx 4776 are attractive low-cobalt hardfacing alloys if the applied loads are less than or equal to103 MPa. The cobalt-base alloys performed best at high loads (207 MPa). Ongoing laboratory studies address the development and evaluation of cobalt-free iron-base hardfacing alloys and seek to improve the wear resistance of cobalt-base alloys by using lasers. Reducing cobalt impurity levels in core components that are periodically discharged should also help reduce radiation fields and disposal costs.

  13. Pressurized water reactor fuel crud and corrosion modeling

    NASA Astrophysics Data System (ADS)

    Deshon, Jeff; Hussey, Dennis; Kendrick, Brian; McGurk, John; Secker, Jeff; Short, Michael

    2011-08-01

    Pressurized water reactors circulate high-temperature water that slowly corrodes Inconel and stainless steel system surfaces, and the nickel/iron based corrosion products deposit in regions of the fuel where sub-cooled nucleate boiling occurs. The deposited corrosion products, called `crud', can have an adverse impact on fuel performance. Boron can concentrate within the crud in the boiling regions of the fuel leading to a phenomenon known as axial offset anomaly (AOA). In rare cases, fuel clad integrity can be compromised because of crud-induced localized corrosion (CILC) of the zirconium-based alloy. Westinghouse and the Electric Power Research Institute have committed to understanding the crud transport process and develop a risk assessment software tool called boron-induced offset anomaly (BOA) to avoid AOA and CILC. This paper reviews the history of the BOA model development and new efforts to develop a micro-scale model called MAMBA for use in the Consortium for Advanced Light Water Reactor Simulation (CASL) program.

  14. Boiling-Water Reactor internals aging degradation study. Phase 1

    SciTech Connect

    Luk, K.H.

    1993-09-01

    This report documents the results of an aging assessment study for boiling water reactor (BWR) internals. Major stressors for BWR internals are related to unsteady hydrodynamic forces generated by the primary coolant flow in the reactor vessel. Welding and cold-working, dissolved oxygen and impurities in the coolant, applied loads and exposures to fast neutron fluxes are other important stressors. Based on results of a component failure information survey, stress corrosion cracking (SCC) and fatigue are identified as the two major aging-related degradation mechanisms for BWR internals. Significant reported failures include SCC in jet-pump holddown beams, in-core neutron flux monitor dry tubes and core spray spargers. Fatigue failures were detected in feedwater spargers. The implementation of a plant Hydrogen Water Chemistry (HWC) program is considered as a promising method for controlling SCC problems in BWR. More operating data are needed to evaluate its effectiveness for internal components. Long-term fast neutron irradiation effects and high-cycle fatigue in a corrosive environment are uncertainty factors in the aging assessment process. BWR internals are examined by visual inspections and the method is access limited. The presence of a large water gap and an absence of ex-core neutron flux monitors may handicap the use of advanced inspection methods, such as neutron noise vibration measurements, for BWR.

  15. Unusual sources of aluminium and heavy metals in potable waters.

    PubMed

    Fuge, R; Pearce, N J; Perkins, W T

    1992-04-01

    Aluminium in water supplies derives from natural sources and from the use of Al2(SO4)3 in water treatment. Heavy metals such as Pb, Cu, Zn and Cd can be added to water from pipework and solder. However, it is apparent that AI and other metals in potable waters can derive from deposits on pipe walls which can be subsequently mobilised when the supply and/or treatment process is changed. Concentrations of Al in domestic supply water of the Llanbrynmair area have been shown to increase from 1 μg to 50 μg L(-1) during its 18 km journey along the water main. Similarly, Pb concentrations in a public building in the Aberystwyth area are found to be extremely elevated due to the metal's mobilisation from encrustations occurring on the copper pipework.

  16. Generic safety insights for inspection of boiling water reactors

    SciTech Connect

    Higgins, J.C.; Taylor, J.H.; Fresco, A.N.; Hillman, B.M.

    1987-01-01

    As the number of operating nuclear power plants (NPPs) increases, safety inspection has increased in importance. Over the last 2 yr, probabilistic risk assessment (PRA) techniques have been developed to aid in the inspection process. Broad interest in generic PRA-based methods has arisen in the past year, since only approx. 25% of the US nuclear power plants have completed PRAs, and also, inspectors want PRA-based tools for these plants. This paper describes the Brookhaven National Lab. program to develop generic boiling water reactor (BWR) PRA-based inspection insights or inspection guidance designed to be applied to plants without PRAs.

  17. Structural Integrity of Water Reactor Pressure Boundary Components.

    DTIC Science & Technology

    1980-08-01

    tests of reference steels of the NRC light water reactor, pressure vessel irradiation dosimetry program. SECURITY CLAS5IICATION 0PHiS PA6GMbn" Dfat ...multiple specimen R- curve approach; NRL emphasis was on the SSC procedure as it is being developed for hot- cell testing of irradiated materials. MULTIPLE...a second autoclave, capable of testing 50 or 100 mm (2T or 4T) thick CT or WOL specimens, was installed in a hot cell and a test was started on 2T-CT

  18. Advanced fuels for plutonium management in pressurized water reactors

    NASA Astrophysics Data System (ADS)

    Vasile, A.; Dufour, Ph; Golfier, H.; Grouiller, J. P.; Guillet, J. L.; Poinot, Ch; Youinou, G.; Zaetta, A.

    2003-06-01

    Several fuel concepts are under investigation at CEA with the aim of manage plutonium inventories in pressurized water reactors. This options range from the use of mature technologies like MOX adapted in the case of MOX-EUS (enriched uranium support) and COmbustible Recyclage A ILot (CORAIL) assemblies to more innovative technologies using IMF like DUPLEX and advanced plutonium assembly (APA). The plutonium burning performances reported to the electrical production go from 7 to 60 kg (TW h) -1. More detailed analysis covering economic, sustainability, reliability and safety aspects and their integration in the whole fuel cycle would allow identifying the best candidate.

  19. 76 FR 52994 - Application for a License To Export Heavy Water

    Federal Register 2010, 2011, 2012, 2013, 2014

    2011-08-24

    ... COMMISSION Application for a License To Export Heavy Water Pursuant to 10 CFR 110.70 (b) ``Public Notice of... end-use for China. (D2O--heavy (liters). producing an active water). pharmaceutical ingredient known as CTP-499, which incorporates heavy water as the source of deuterium to achieve the...

  20. Reactor physics behavior of transuranic-bearing TRISO-particle fuel in a pressurized water reactor

    SciTech Connect

    Pope, M. A.; Sen, R. S.; Ougouag, A. M.; Youinou, G.; Boer, B.

    2012-07-01

    Calculations have been performed to assess the neutronic behavior of pins of Fully-Ceramic Micro-encapsulated (FCM) fuel in otherwise-conventional Pressurized Water Reactor (PWR) fuel pins. The FCM fuel contains transuranic (TRU) - only oxide fuel in tri-isotropic (TRISO) particles with the TRU loading coming from the spent fuel of a conventional LWR after 5 years of cooling. Use of the TRISO particle fuel would provide an additional barrier to fission product release in the event of cladding failure. Depletion calculations were performed to evaluate reactivity-limited burnup of the TRU-only FCM fuel. These calculations showed that due to relatively little space available for fuel, the achievable burnup with these pins alone is quite small. Various reactivity parameters were also evaluated at each burnup step including moderator temperature coefficient (MTC), Doppler, and soluble boron worth. These were compared to reference UO{sub 2} and MOX unit cells. The TRU-only FCM fuel exhibits degraded MTC and Doppler coefficients relative to UO{sub 2} and MOX. Also, the reactivity effects of coolant voiding suggest that the behavior of this fuel would be similar to a MOX fuel of very high plutonium fraction, which are known to have positive void reactivity. In general, loading of TRU-only FCM fuel into an assembly without significant quantities of uranium presents challenges to the reactor design. However, if such FCM fuel pins are included in a heterogeneous assembly alongside LEU fuel pins, the overall reactivity behavior would be dominated by the uranium pins while attractive TRU destruction performance levels in the TRU-only FCM fuel pins is retained. From this work, it is concluded that use of heterogeneous assemblies such as these appears feasible from a preliminary reactor physics standpoint. (authors)

  1. Reactor Physics Behavior of Transuranic-Bearing TRISO-Particle Fuel in a Pressurized Water Reactor

    SciTech Connect

    Michael A. Pope; R. Sonat Sen; Abderrafi M. Ougouag; Gilles Youinou; Brian Boer

    2012-04-01

    Calculations have been performed to assess the neutronic behavior of pins of Fully-Ceramic Micro-encapsulated (FCM) fuel in otherwise-conventional Pressurized Water Reactor (PWR) fuel pins. The FCM fuel contains transuranic (TRU)-only oxide fuel in tri-isotropic (TRISO) particles with the TRU loading coming from the spent fuel of a conventional LWR after 5 years of cooling. Use of the TRISO particle fuel would provide an additional barrier to fission product release in the event of cladding failure. Depletion calculations were performed to evaluate reactivity-limited burnup of the TRU-only FCM fuel. These calculations showed that due to relatively little space available for fuel, the achievable burnup with these pins alone is quite small. Various reactivity parameters were also evaluated at each burnup step including moderator temperature coefficient (MTC), Doppler, and soluble boron worth. These were compared to reference UO{sub 2} and MOX unit cells. The TRU-only FCM fuel exhibits degraded MTC and Doppler coefficients relative to UO{sub 2} and MOX. Also, the reactivity effects of coolant voiding suggest that the behavior of this fuel would be similar to a MOX fuel of very high plutonium fraction, which are known to have positive void reactivity. In general, loading of TRU-only FCM fuel into an assembly without significant quantities of uranium presents challenges to the reactor design. However, if such FCM fuel pins are included in a heterogeneous assembly alongside LEU fuel pins, the overall reactivity behavior would be dominated by the uranium pins while attractive TRU destruction performance levels in the TRU-only FCM fuel pins is. From this work, it is concluded that use of heterogeneous assemblies such as these appears feasible from a preliminary reactor physics standpoint.

  2. Light Water Reactor Sustainability Constellation Pilot Project FY11 Summary Report

    SciTech Connect

    R. Johansen

    2011-09-01

    Summary report for Fiscal Year 2011 activities associated with the Constellation Pilot Project. The project is a joint effor between Constellation Nuclear Energy Group (CENG), EPRI, and the DOE Light Water Reactor Sustainability Program. The project utilizes two CENG reactor stations: R.E. Ginna and Nine Point Unit 1. Included in the report are activities associate with reactor internals and concrete containments.

  3. Design of Recycle PWR with Heavy Water Moderation

    SciTech Connect

    Hibi, K.; Uchita, M.

    2002-07-01

    This study shows the conceptual plant design of the recycle PWR (RPWR), which is an innovative MOX-PWR with breeding ratios around 1.1 moderated by heavy water. Most of the plant systems of RPWR can employ the systems of PWRs. RPWR has no acid boron systems and has a small tritium removal system. The construction and operation costs are similar to the current PWRs. While, heavy water cost will be decreased drastically with up-to-date producing methods. The reliability for the plant systems of RPWR is high and R and D cost for realizing RPWR is very low because the core design of RPWR is fundamentally based on the current PWR technology. (authors)

  4. High Temperature Water Heat Pipes Radiator for a Brayton Space Reactor Power System

    SciTech Connect

    El-Genk, Mohamed S.; Tournier, Jean-Michel

    2006-01-20

    A high temperature water heat pipes radiator design is developed for a space power system with a sectored gas-cooled reactor and three Closed Brayton Cycle (CBC) engines, for avoidance of single point failures in reactor cooling and energy conversion and rejection. The CBC engines operate at turbine inlet and exit temperatures of 1144 K and 952 K. They have a net efficiency of 19.4% and each provides 30.5 kWe of net electrical power to the load. A He-Xe gas mixture serves as the turbine working fluid and cools the reactor core, entering at 904 K and exiting at 1149 K. Each CBC loop is coupled to a reactor sector, which is neutronically and thermally coupled, but hydraulically decoupled to the other two sectors, and to a NaK-78 secondary loop with two water heat pipes radiator panels. The segmented panels each consist of a forward fixed segment and two rear deployable segments, operating hydraulically in parallel. The deployed radiator has an effective surface area of 203 m2, and when the rear segments are folded, the stowed power system fits in the launch bay of the DELTA-IV Heavy launch vehicle. For enhanced reliability, the water heat pipes operate below 50% of their wicking limit; the sonic limit is not a concern because of the water, high vapor pressure at the temperatures of interest (384 - 491 K). The rejected power by the radiator peaks when the ratio of the lengths of evaporator sections of the longest and shortest heat pipes is the same as that of the major and minor widths of the segments. The shortest and hottest heat pipes in the rear segments operate at 491 K and 2.24 MPa, and each rejects 154 W. The longest heat pipes operate cooler (427 K and 0.52 MPa) and because they are 69% longer, reject more power (200 W each). The longest and hottest heat pipes in the forward segments reject the largest power (320 W each) while operating at {approx} 46% of capillary limit. The vapor temperature and pressure in these heat pipes are 485 K and 1.97 MPa. By

  5. High Temperature Water Heat Pipes Radiator for a Brayton Space Reactor Power System

    NASA Astrophysics Data System (ADS)

    El-Genk, Mohamed S.; Tournier, Jean-Michel

    2006-01-01

    A high temperature water heat pipes radiator design is developed for a space power system with a sectored gas-cooled reactor and three Closed Brayton Cycle (CBC) engines, for avoidance of single point failures in reactor cooling and energy conversion and rejection. The CBC engines operate at turbine inlet and exit temperatures of 1144 K and 952 K. They have a net efficiency of 19.4% and each provides 30.5 kWe of net electrical power to the load. A He-Xe gas mixture serves as the turbine working fluid and cools the reactor core, entering at 904 K and exiting at 1149 K. Each CBC loop is coupled to a reactor sector, which is neutronically and thermally coupled, but hydraulically decoupled to the other two sectors, and to a NaK-78 secondary loop with two water heat pipes radiator panels. The segmented panels each consist of a forward fixed segment and two rear deployable segments, operating hydraulically in parallel. The deployed radiator has an effective surface area of 203 m2, and when the rear segments are folded, the stowed power system fits in the launch bay of the DELTA-IV Heavy launch vehicle. For enhanced reliability, the water heat pipes operate below 50% of their wicking limit; the sonic limit is not a concern because of the water, high vapor pressure at the temperatures of interest (384 - 491 K). The rejected power by the radiator peaks when the ratio of the lengths of evaporator sections of the longest and shortest heat pipes is the same as that of the major and minor widths of the segments. The shortest and hottest heat pipes in the rear segments operate at 491 K and 2.24 MPa, and each rejects 154 W. The longest heat pipes operate cooler (427 K and 0.52 MPa) and because they are 69% longer, reject more power (200 W each). The longest and hottest heat pipes in the forward segments reject the largest power (320 W each) while operating at ~ 46% of capillary limit. The vapor temperature and pressure in these heat pipes are 485 K and 1.97 MPa. By contrast, the

  6. Copper removal using a heavy-metal resistant microbial consortium in a fixed-bed reactor.

    PubMed

    Carpio, Isis E Mejias; Machado-Santelli, Glaucia; Sakata, Solange Kazumi; Ferreira Filho, Sidney Seckler; Rodrigues, Debora Frigi

    2014-10-01

    A heavy-metal resistant bacterial consortium was obtained from a contaminated river in São Paulo, Brazil and utilized for the design of a fixed-bed column for the removal of copper. Prior to the design of the fixed-bed bioreactor, the copper removal capacity by the live consortium and the effects of copper in the consortium biofilm formation were investigated. The Langmuir model indicated that the sorption capacity of the consortium for copper was 450.0 mg/g dry cells. The biosorption of copper into the microbial biomass was attributed to carboxyl and hydroxyl groups present in the microbial biomass. The effect of copper in planktonic cells to form biofilm under copper rich conditions was investigated with confocal microscopy. The results revealed that biofilm formed after 72 h exposure to copper presented a reduced thickness by 57% when compared to the control; however 84% of the total cells were still alive. The fixed-bed bioreactor was set up by growing the consortium biofilm on granular activated carbon (GAC) and analyzed for copper removal. The biofilm-GAC (BGAC) column retained 45% of the copper mass present in the influent, as opposed to 17% in the control column that contained GAC only. These findings suggest that native microbial communities in sites contaminated with heavy metals can be immobilized in fixed-bed bioreactors and used to treat metal contaminated water.

  7. Continuous-flow solar UVB disinfection reactor for drinking water.

    PubMed

    Mbonimpa, Eric Gentil; Vadheim, Bryan; Blatchley, Ernest R

    2012-05-01

    Access to safe, reliable sources of drinking water is a long-standing problem among people in developing countries. Sustainable solutions to these problems often involve point-of-use or community-scale water treatment systems that rely on locally-available resources and expertise. This philosophy was used in the development of a continuous-flow, solar UVB disinfection system. Numerical modeling of solar UVB spectral irradiance was used to define temporal variations in spectral irradiance at several geographically-distinct locations. The results of these simulations indicated that a solar UVB system would benefit from incorporation of a device to amplify ambient UVB fluence rate. A compound parabolic collector (CPC) was selected for this purpose. Design of the CPC was based on numerical simulations that accounted for the shape of the collector and reflectance. Based on these simulations, a prototype CPC was constructed using materials that would be available and inexpensive in many developing countries. A UVB-transparent pipe was positioned in the focal area of the CPC; water was pumped through the pipe to allow exposure of waterborne microbes to germicidal solar UVB radiation. The system was demonstrated to be effective for inactivation of Escherichia coli, and DNA-weighted UV dose was shown to govern reactor performance. The design of the reactor is expected to scale linearly, and improvements in process performance (relative to results from the prototype) can be expected by use of larger CPC geometry, inclusion of better reflective materials, and application in areas with greater ambient solar UV spectral irradiance than the location of the prototype tests. The system is expected to have application for water treatment among communities in (developing) countries in near-equatorial and tropical locations. It may also have application for disaster relief or military field operations, as well as in water treatment in areas of developed countries that receive

  8. Summary of Research on Light Water Reactor Improvement Concepts

    SciTech Connect

    Mowery, Alfred L

    2002-12-15

    The Arms Control and Disarmament Agency of the U.S. Department of State instituted a study aimed at improving the light water reactor (LWR) fuel consumption efficiency as an alternative to fuel recycle in the late 1970s. Comparison of the neutron balance tables of an LWR (1982 design) and an 'advanced' Canada deuterium uranium (CANDU) reactor explained that the relatively low fuel efficiency of the LWR was not primarily a consequence of water moderator absorptions. Rather, the comparatively low LWR fuel efficiency resulted from its use of poison to hold down startup reactivity together with other neutron losses. The research showed that each neutron saved could reduce fuel consumption by about 5%. In a typical LWR some 5 neutrons (out of 100) were absorbed in control poisons over a cycle. There are even more parasitic and leakage neutron absorptions. The objective of the research was to find ways to minimize control, parasitic, and other neutron losses aimed at improved LWR fuel consumption. Further research developed the concept of 'putting neutrons in the bank' in {sup 238}U early in life and 'drawing them out of the bank' late in life by burning the {sup 239}Pu produced. Conceptual designs were explored that could both control the reactor and substantially improve fuel efficiency and minimize separative work requirements.The U.S. Department of Energy augmented its high burnup fuel program based on the research in the late 1970s. As a result of the success of this program, fuel burnup in U.S. LWRs has almost doubled in the intervening two decades.

  9. DEGRADATION EVALUATION OF HEAVY WATER DRUMS AND TANKS

    SciTech Connect

    Mickalonis, J.; Vormelker, P.

    2009-07-31

    Heavy water with varying chemistries is currently being stored in over 6700 drums in L- and K-areas and in seven tanks in L-, K-, and C-areas. A detailed evaluation of the potential degradation of the drums and tanks, specific to their design and service conditions, has been performed to support the demonstration of their integrity throughout the desired storage period. The 55-gallon drums are of several designs with Type 304 stainless steel as the material of construction. The tanks have capacities ranging from 8000 to 45600 gallons and are made of Type 304 stainless steel. The drums and tanks were designed and fabricated to national regulations, codes and standards per procurement specifications for the Savannah River Site. The drums have had approximately 25 leakage failures over their 50+ years of use with the last drum failure occurring in 2003. The tanks have experienced no leaks to date. The failures in the drums have occurred principally near the bottom weld, which attaches the bottom to the drum sidewall. Failures have occurred by pitting, crevice and stress corrosion cracking and are attributable, in part, to the presence of chloride ions in the heavy water. Probable degradation mechanisms for the continued storage of heavy water were evaluated that could lead to future failures in the drum or tanks. This evaluation will be used to support establishment of an inspection plan which will include susceptible locations, methods, and frequencies for the drums and tanks to avoid future leakage failures.

  10. DOE/NNSA perspective safeguard by design: GEN III/III+ light water reactors and beyond

    SciTech Connect

    Pan, Paul Y

    2010-12-10

    An overview of key issues relevant to safeguards by design (SBD) for GEN III/IV nuclear reactors is provided. Lessons learned from construction of typical GEN III+ water reactors with respect to SBD are highlighted. Details of SBD for safeguards guidance development for GEN III/III+ light water reactors are developed and reported. This paper also identifies technical challenges to extend SBD including proliferation resistance methodologies to other GEN III/III+ reactors (except HWRs) and GEN IV reactors because of their immaturity in designs.

  11. An Interactive Code for a Pressurized Water Reactor Incorporating Temperature and Xenon Feedback.

    DTIC Science & Technology

    1980-06-01

    feedback effects , as well as an automatic reactor protection and average reactor coolant temperature control system. The thermal response of the model plant ...of a pressurized water reactor power plant . The reactivity feedback effects of Xenon-135 poisoning and moderator temperature are incorporated into the...developed, using the applicable plant parameters from Shippingport Atomic Power Station. The point reactor kinetics equations for one delayed neutron

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

    SciTech Connect

    Hellesen, C.; Grape, S.; Haakanson, A.; Jacobson Svaerd, S.; Jansson, P.

    2013-07-01

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

  13. Sustained Recycle in Light Water and Sodium-Cooled Reactors

    SciTech Connect

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

    2010-11-01

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

  14. Development of online, continuous heavy metals detection and monitoring sensors based on microfluidic plasma reactors

    NASA Astrophysics Data System (ADS)

    Abdul-Majeed, Wameath Sh

    This research is dedicated to develop a fully integrated system for heavy metals determination in water samples based on micro fluidic plasma atomizers. Several configurations of dielectric barrier discharge (DBD) atomizer are designed, fabricated and tested toward this target. Finally, a combination of annular and rectangular DBD atomizers has been utilized to develop a scheme for heavy metals determination. The present thesis has combined both theoretical and experimental investigations to fulfil the requirements. Several mathematical studies are implemented to explore the optimal design parameters for best system performance. On the other hand, expanded experimental explorations are conducted to assess the proposed operational approaches. The experiments were designed according to a central composite rotatable design; hence, an empirical model has been produced for each studied case. Moreover, several statistical approaches are adopted to analyse the system performance and to deduce the optimal operational parameters.. The introduction of the examined analyte to the plasma atomizer has been achieved by applying chemical schemes, where the element in the sample has been derivitized by using different kinds of reducing agents to produce vapour species (e.g. hydrides) for a group of nine elements examined in this research individually and simultaneously. Moreover, other derivatization schemes based on photochemical vapour generation assisted by ultrasound irradiation are also investigated. Generally speaking, the detection limits achieved in this research for the examined set of elements (by applying hydroborate scheme) are found to be acceptable in accordance with the standard limits in drinking water. The results of copper compared with the data from other technologies in the literature, showed a competitive detection limit obtained from applying the developed scheme, with an advantage of conducting simultaneous, fully automated, insitu, online- real time

  15. Salt deposition studies in a supercritical water oxidation reactor

    SciTech Connect

    LaJeunesse, C.A.; Rice, S.F.; Hanush, R.G.; Aiken, J.D.

    1993-10-01

    Supercritical water oxidation (SCWO), a method for destroying aqueous organic waste, is a relatively new technology discovered about fifteen years ago. SCWO occurs at moderate temperatures and pressures where the ability of water to dissolve hydrocarbons and other organic chemicals is greatly enhanced. Depending on the feed stream and residence time, the dissolved organic waste reacts with an oxidizer to produce innocuous combustion products. However, oxidation of organic material containing heteroatoms, such as sulfur or phosphorous, forms sulfuric or phosphoric acid in the absence of metal ions. In situ neutralization with sodium hydroxide then forms salts that are insoluble at supercritical conditions. These salts deposit in the reactor affecting the processing of the organic material. To design a system that can accommodate the formation of these salts, it is important to understand the deposition process quantitatively. This paper is an interim report on an experimental program designed to understand the salt deposition phenomena.

  16. Vibrational spectra of light and heavy water with application to neutron cross section calculations

    SciTech Connect

    Damian, J. I. Marquez; Granada, J. R.; Malaspina, D. C.

    2013-07-14

    The design of nuclear reactors and neutron moderators require a good representation of the interaction of low energy (E < 1 eV) neutrons with hydrogen and deuterium containing materials. These models are based on the dynamics of the material, represented by its vibrational spectrum. In this paper, we show calculations of the frequency spectrum for light and heavy water at room temperature using two flexible point charge potentials: SPC-MPG and TIP4P/2005f. The results are compared with experimental measurements, with emphasis on inelastic neutron scattering data. Finally, the resulting spectra are applied to calculation of neutron scattering cross sections for these materials, which were found to be a significant improvement over library data.

  17. Mid-IR absorption sensing of heavy water using a silicon-on-sapphire waveguide.

    PubMed

    Singh, Neetesh; Casas-Bedoya, Alvaro; Hudson, Darren D; Read, Andrew; Mägi, Eric; Eggleton, Benjamin J

    2016-12-15

    We demonstrate a compact silicon-on-sapphire (SOS) strip waveguide sensor for mid-IR absorption spectroscopy. This device can be used for gas and liquid sensing, especially to detect chemically similar molecules and precisely characterize extremely absorptive liquids that are difficult to detect by conventional infrared transmission techniques. We reliably measure concentrations up to 0.25% of heavy water (D2O) in a D2O-H2O mixture at its maximum absorption band at around 4 μm. This complementary metal-oxide-semiconductor (CMOS) compatible SOS D2O sensor is promising for applications such as measuring body fat content or detection of coolant leakage in nuclear reactors.

  18. Light Water Reactor Sustainability Program Integrated Program Plan

    SciTech Connect

    Kathryn McCarthy; Jeremy Busby; Bruce Hallbert; Shannon Bragg-Sitton; Curtis Smith; Cathy Barnard

    2013-04-01

    Nuclear power has safely, reliably, and economically contributed almost 20% of electrical generation in the United States over the past two decades. It remains the single largest contributor (more than 70%) of non-greenhouse-gas-emitting electric power generation in the United States. Domestic demand for electrical energy is expected to experience a 31% growth from 2009 to 2035. At the same time, most of the currently operating nuclear power plants will begin reaching the end of their initial 20-year extension to their original 40-year operating license for a total of 60 years of operation. Figure E-1 shows projected nuclear energy contribution to the domestic generating capacity. If current operating nuclear power plants do not operate beyond 60 years, the total fraction of generated electrical energy from nuclear power will begin to decline—even with the expected addition of new nuclear generating capacity. The oldest commercial plants in the United States reached their 40th anniversary in 2009. The U.S. Department of Energy Office of Nuclear Energy’s Research and Development Roadmap (Nuclear Energy Roadmap) organizes its activities around four objectives that ensure nuclear energy remains a compelling and viable energy option for the United States. The four objectives are as follows: (1) develop technologies and other solutions that can improve the reliability, sustain the safety, and extend the life of the current reactors; (2) develop improvements in the affordability of new reactors to enable nuclear energy to help meet the Administration’s energy security and climate change goals; (3) develop sustainable nuclear fuel cycles; and (4) understand and minimize the risks of nuclear proliferation and terrorism. The Light Water Reactor Sustainability (LWRS) Program is the primary programmatic activity that addresses Objective 1. This document summarizes the LWRS Program’s plans.

  19. Light Water Reactor Sustainability Program Integrated Program Plan

    SciTech Connect

    McCarthy, Kathryn A.; Busby, Jeremy; Hallbert, Bruce; Bragg-Sitton, Shannon; Smith, Curtis; Barnard, Cathy

    2014-04-01

    Nuclear power has safely, reliably, and economically contributed almost 20% of electrical generation in the United States over the past two decades. It remains the single largest contributor (more than 70%) of non-greenhouse-gas-emitting electric power generation in the United States. Domestic demand for electrical energy is expected to experience a 31% growth from 2009 to 2035. At the same time, most of the currently operating nuclear power plants will begin reaching the end of their initial 20-year extension to their original 40-year operating license for a total of 60 years of operation. Figure E-1 shows projected nuclear energy contribution to the domestic generating capacity. If current operating nuclear power plants do not operate beyond 60 years, the total fraction of generated electrical energy from nuclear power will begin to decline—even with the expected addition of new nuclear generating capacity. The oldest commercial plants in the United States reached their 40th anniversary in 2009. The U.S. Department of Energy Office of Nuclear Energy’s Research and Development Roadmap (Nuclear Energy Roadmap) organizes its activities around four objectives that ensure nuclear energy remains a compelling and viable energy option for the United States. The four objectives are as follows: (1) develop technologies and other solutions that can improve the reliability, sustain the safety, and extend the life of the current reactors; (2) develop improvements in the affordability of new reactors to enable nuclear energy to help meet the Administration’s energy security and climate change goals; (3) develop sustainable nuclear fuel cycles; and (4) understand and minimize the risks of nuclear proliferation and terrorism. The Light Water Reactor Sustainability (LWRS) Program is the primary programmatic activity that addresses Objective 1. This document summarizes the LWRS Program’s plans.

  20. Revised accident source terms for light-water reactors

    SciTech Connect

    Soffer, L.

    1995-02-01

    This paper presents revised accident source terms for light-water reactors incorporating the severe accident research insights gained in this area over the last 15 years. Current LWR reactor accident source terms used for licensing date from 1962 and are contained in Regulatory Guides 1.3 and 1.4. These specify that 100% of the core inventory of noble gases and 25% of the iodine fission products are assumed to be instantaneously available for release from the containment. The chemical form of the iodine fission products is also assumed to be predominantly elemental iodine. These assumptions have strongly affected present nuclear air cleaning requirements by emphasizing rapid actuation of spray systems and filtration systems optimized to retain elemental iodine. A proposed revision of reactor accident source terms and some im implications for nuclear air cleaning requirements was presented at the 22nd DOE/NRC Nuclear Air Cleaning Conference. A draft report was issued by the NRC for comment in July 1992. Extensive comments were received, with the most significant comments involving (a) release fractions for both volatile and non-volatile species in the early in-vessel release phase, (b) gap release fractions of the noble gases, iodine and cesium, and (c) the timing and duration for the release phases. The final source term report is expected to be issued in late 1994. Although the revised source terms are intended primarily for future plants, current nuclear power plants may request use of revised accident source term insights as well in licensing. This paper emphasizes additional information obtained since the 22nd Conference, including studies on fission product removal mechanisms, results obtained from improved severe accident code calculations and resolution of major comments, and their impact upon the revised accident source terms. Revised accident source terms for both BWRS and PWRS are presented.

  1. Light Water Reactor Sustainability Program Integrated Program Plan

    SciTech Connect

    George Griffith; Robert Youngblood; Jeremy Busby; Bruce Hallbert; Cathy Barnard; Kathryn McCarthy

    2012-01-01

    Nuclear power has safely, reliably, and economically contributed almost 20% of electrical generation in the United States over the past two decades. It remains the single largest contributor (more than 70%) of non-greenhouse-gas-emitting electric power generation in the United States. Domestic demand for electrical energy is expected to experience a 31% growth from 2009 to 2035. At the same time, most of the currently operating nuclear power plants will begin reaching the end of their initial 20-year extension to their original 40-year operating license for a total of 60 years of operation. Figure E-1 shows projected nuclear energy contribution to the domestic generating capacity. If current operating nuclear power plants do not operate beyond 60 years, the total fraction of generated electrical energy from nuclear power will begin to decline - even with the expected addition of new nuclear generating capacity. The oldest commercial plants in the United States reached their 40th anniversary in 2009. The U.S. Department of Energy Office of Nuclear Energy's Research and Development Roadmap (Nuclear Energy Roadmap) organizes its activities around four objectives that ensure nuclear energy remains a compelling and viable energy option for the United States. The four objectives are as follows: (1) develop technologies and other solutions that can improve the reliability, sustain the safety, and extend the life of the current reactors; (2) develop improvements in the affordability of new reactors to enable nuclear energy to help meet the Administration's energy security and climate change goals; (3) develop sustainable nuclear fuel cycles; and (4) understand and minimize the risks of nuclear proliferation and terrorism. The Light Water Reactor Sustainability (LWRS) Program is the primary programmatic activity that addresses Objective 1. This document summarizes the LWRS Program's plans.

  2. Effects of Water Radiolysis in Water Cooled Reactors - Nuclear Energy Research Initiative (NERI) Program

    SciTech Connect

    S. M. Pimblott

    2000-10-01

    OAK B188 Quarterly Progress Report on NERI Proposal No.99-0010 for the Development of an Experiment and Calculation Based Model to Describe the Effects of Radiation on Non-standard Aqueous Systems Like Those Encountered in the Advanced Light Water Reactor

  3. Catalytic membrane reactor for water and wastewater treatment

    NASA Astrophysics Data System (ADS)

    Heng, Samuel

    A double membrane reactor was fabricated and assessed for continuous treatment of water containing organic contaminants by ozonation. This innovative reactor consisted of a zeolite membrane prepared on the inner surface of a porous a-alumina support, which served as water selective extractor and active contactor, and a porous stainless membrane which was the ozone gas diffuser. The coupling of membrane separation and chemical oxidation was found to be highly beneficial to both processes. The total organic carbon (TOC) removal rate at the retentate was enhanced by up to 2.2 times, as compared to membrane ozonation. Simultaneously, clean water (< 2 mg C.L-1 ) was consistently produced on the permeate side, using a feed solution containing up to 1000 mg C.L-1, while the retentate was concentrated and treated. Most significantly, the addition of an adsorbing material, as a bed or a coated layer, onto the pores of the membrane support, was shown to further enhance TOC degradation, permeated TOC concentration, permeate flux, and moreover, ozone yield. The achievements of this project included: (1) The development of a novel low-temperature zeolite membrane activation method that generates consistently high quality membranes (i.e. high reproducibility and fewer defects). (2) The demonstration that gamma-alumina and gamma-alumina supported catalysts do not have significant activity and that the TOC removal enhancement usually observed during catalytic ozonation was due primarily to the contribution of adsorption and metal leaching. Thermogravimetric analysis (TGA) and elemental analysis (EA) of the spent catalyst showed that, during catalytic ozonation, oxygenated by-products of increased adsorbability were concentrated onto the gamma-alumina contactor, and were subsequently degraded. (3) The development of a method for coating high surface area gamma-alumina layers onto the grains of zeolite membrane support used as the active membrane contactor.

  4. Light Water Reactor Sustainability Program Reactor Safety Technologies Pathway Technical Program Plan

    SciTech Connect

    Corradini, M. L.; Peko, D.; Farmer, M.; Rempe, J.; Humrickhouse, P.; O'Brien, J.; Robb, K.; Gauntt, R.; Osborn, D.

    2016-06-01

    “end user” of the results from this DOE-sponsored work. The response to the Fukushima accident has been global, and there is a continuing multinational interest in collaborations to better quantify accident consequences and to incorporate lessons learned from the accident. DOE will continue to seek opportunities to facilitate collaborations that are of value to the U.S. industry, particularly where the collaboration provides access to vital data from the accident or otherwise supports or leverages other important R&D work. The purpose of the Reactor Safety Technology R&D is to improve understanding of beyond design basis events and reduce uncertainty in severe accident progression, phenomenology, and outcomes using existing analytical codes and information gleaned from severe accidents, in particular the Fukushima Daiichi events. This information will be used to aid in developing mitigating strategies and improving severe accident management guidelines for the current light water reactor fleet.

  5. Light Water Reactor Sustainability Program: Reactor Safety Technologies Pathway Technical Program Plan

    SciTech Connect

    Corradini, M. L.

    2015-06-01

    “end user” of the results from this DOE-sponsored work. The response to the Fukushima accident has been global, and there is a continuing multinational interest in collaborations to better quantify accident consequences and to incorporate lessons learned from the accident. DOE will continue to seek opportunities to facilitate collaborations that are of value to the U.S. industry, particularly where the collaboration provides access to vital data from the accident or otherwise supports or leverages other important R&D work. The purpose of the Reactor Safety Technology R&D is to improve understanding of beyond design basis events and reduce uncertainty in severe accident progression, phenomenology, and outcomes using existing analytical codes and information gleaned from severe accidents, in particular the Fukushima Daiichi events. This information will be used to aid in developing mitigating strategies and improving severe accident management guidelines for the current light water reactor fleet.

  6. [Heavy metals in water of the Skikda Bay].

    PubMed

    Kehal, M; Mennour, A; Reinert, L; Fuzellier, H

    2004-09-01

    The region of Skikda is one of the most important industrial poles of Algeria. The aim of the study is a qualitative and quantitative evaluation of the pollution by heavy metals of the marine water of the bay. The pollutants investigated are lead, cadmium and mercury because of their toxicity. The study is concerned mainly with the spatiotemporal evolution of the pollution on the extent of the bay. Concentrations of heavy metals metals vary from 4 microg l(-1) to 55 microg l(-1) for lead, 1 microg l(-1) to 17 microg l(-1) for cadmium and 0,1 to 1,1 microg l(-1) for mercury, which indicates a beginning of pollution of the site. Only small variation of the contents have been noted in a second investigation carried out one decade after the first one.

  7. New evaluation of thermal neutron scattering libraries for light and heavy water

    NASA Astrophysics Data System (ADS)

    Marquez Damian, Jose Ignacio; Granada, Jose Rolando; Cantargi, Florencia; Roubtsov, Danila

    2017-09-01

    In order to improve the design and safety of thermal nuclear reactors and for verification of criticality safety conditions on systems with significant amount of fissile materials and water, it is necessary to perform high-precision neutron transport calculations and estimate uncertainties of the results. These calculations are based on neutron interaction data distributed in evaluated nuclear data libraries. To improve the evaluations of thermal scattering sub-libraries, we developed a set of thermal neutron scattering cross sections (scattering kernels) for hydrogen bound in light water, and deuterium and oxygen bound in heavy water, in the ENDF-6 format from room temperature up to the critical temperatures of molecular liquids. The new evaluations were generated and processable with NJOY99 and also with NJOY-2012 with minor modifications (updates), and with the new version of NJOY-2016. The new TSL libraries are based on molecular dynamics simulations with GROMACS and recent experimental data, and result in an improvement of the calculation of single neutron scattering quantities. In this work, we discuss the importance of taking into account self-diffusion in liquids to accurately describe the neutron scattering at low neutron energies (quasi-elastic peak problem). To improve modeling of heavy water, it is important to take into account temperature-dependent static structure factors and apply Sköld approximation to the coherent inelastic components of the scattering matrix. The usage of the new set of scattering matrices and cross-sections improves the calculation of thermal critical systems moderated and/or reflected with light/heavy water obtained from the International Criticality Safety Benchmark Evaluation Project (ICSBEP) handbook. For example, the use of the new thermal scattering library for heavy water, combined with the ROSFOND-2010 evaluation of the cross sections for deuterium, results in an improvement of the C/E ratio in 48 out of 65

  8. Technologies for Upgrading Light Water Reactor Outlet Temperature

    SciTech Connect

    Daniel S. Wendt; Piyush Sabharwall; Vivek Utgikar

    2013-07-01

    Nuclear energy could potentially be utilized in hybrid energy systems to produce synthetic fuels and feedstocks from indigenous carbon sources such as coal and biomass. First generation nuclear hybrid energy system (NHES) technology will most likely be based on conventional light water reactors (LWRs). However, these LWRs provide thermal energy at temperatures of approximately 300°C, while the desired temperatures for many chemical processes are much higher. In order to realize the benefits of nuclear hybrid energy systems with the current LWR reactor fleets, selection and development of a complimentary temperature upgrading technology is necessary. This paper provides an initial assessment of technologies that may be well suited toward LWR outlet temperature upgrading for powering elevated temperature industrial and chemical processes during periods of off-peak power demand. Chemical heat transformers (CHTs) are a technology with the potential to meet LWR temperature upgrading requirements for NHESs. CHTs utilize chemical heat of reaction to change the temperature at which selected heat sources supply or consume thermal energy. CHTs could directly utilize LWR heat output without intermediate mechanical or electrical power conversion operations and the associated thermodynamic losses. CHT thermal characteristics are determined by selection of the chemical working pair and operating conditions. This paper discusses the chemical working pairs applicable to LWR outlet temperature upgrading and the CHT operating conditions required for providing process heat in NHES applications.

  9. Pressurized-water reactor internals aging degradation study. Phase 1

    SciTech Connect

    Luk, K.H.

    1993-09-01

    This report documents the results of a Phase I study on the effects of aging degradations on pr internals. Primary stressers for internals an generated by the primary coolant flow in the they include unsteady hydrodynamic forces and pump-generated pressure pulsations. Other stressors are applied loads, manufacturing processes, impurities in the coolant and exposures to fast neutron fluxes. A survey of reported aging-related failure information indicates that fatigue, stress corrosion cracking (SCC) and mechanical wear are the three major aging-related degradation mechanisms for PWR internals. Significant reported failures include thermal shield flow-induced vibration problems, SCC in guide tube support pins and core support structure bolts, fatigue-induced core baffle water-jet impingement problems and excess wear in flux thimbles. Many of the reported problems have been resolved by accepted engineering practices. Uncertainties remain in the assessment of long-term neutron irradiation effects and environmental factors in high-cycle fatigue failures. Reactor internals are examined by visual inspections and the technique is access limited. Improved inspection methods, especially one with an early failure detection capability, can enhance the safety and efficiency of reactor operations.

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

    SciTech Connect

    Conley, D.A.; Edson, J.L.; Fineman, C.F.

    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.

  11. Thermal-Hydraulic Analysis of Supercritical Pressure Light Water Reactors

    SciTech Connect

    Cheng, X.; Schulenberg, T.; Koshizuka, S.; Oka, Y.; Souyri, A.

    2002-07-01

    In the frame of the European project HPLWR, joined by European research institutions, industrial partners and the University of Tokyo, thermal-hydraulic analysis of supercritical pressure light water reactors has been carried out. A thorough literature survey on heat transfer of supercritical fluids indicates a large deficiency in the prediction of the heat transfer coefficient and the onset of heat transfer deterioration under the reactor condition. A CFD code for analysing the thermal-hydraulic behaviour of supercritical fluids was developed. Numerical results show that the heat transfer coefficient, including the heat transfer deterioration region, can be well predicted using this CFD code, at least for circular tube geometries. Such a CFD code is well suitable for understanding the heat transfer mechanism. Based on the numerical results, a new heat transfer correlation has been proposed. For the thermal-hydraulic design of an HPLWR fuel assembly, the subchannel analysis code STAR-SC has been developed with a high numerical efficiency and a high applicability to different kinds of fuel assembly configurations. The results show clearly that design of a HPLWR fuel assembly is a highly challenging task. At the same time, sub-channel analysis provides some important guidelines for the design of a HPLWR fuel assembly. (authors)

  12. Controlling radiation fields in siemans designed light water reactors

    SciTech Connect

    Riess, R.; Marchl, T.

    1995-03-01

    An essential item for the control of radiation fields is the minimization of the use of satellites in the reactor systems of Light Water Reactors (LWRs). A short description of the qualification of Co-replacement materials will be followed by an illustration of the locations where these materials were implemented in Siemens designed LWRs. Especially experiences in PWRs show the immense influence of reduction of cobalt sources on dose rate buildup. The corrosion and the fatique and wear behavior of the replacement materials has not created concern up to now. A second tool to keep occupational radiation doses at a low level in PWRs is the use of the modified B/Li-chemistry. This is practized in Siemens designed plants by keeping the Li level at a max. value of 2 ppm until it reaches a pH (at 300{degrees}C) of {approximately}7.4. This pH is kept constant until the end of the cycle. The substitution of cobalt base alloys and thus the removal of the Co-59 sources from the system had the largest impact on the radiation levels. Nonetheless, the effectiveness of the coolant chemistry should not be neglected either. Several years of successful operation of PWRs with the replacement materials resulted in an occupational radiation exposure which is below 0.5 man-Sievert/plant and year.

  13. Development of Novel Water-Gas Shift Membrane Reactor

    SciTech Connect

    Ho, W. S. Winston

    2004-12-29

    This report summarizes the objectives, technical barrier, approach, and accomplishments for the development of a novel water-gas-shift (WGS) membrane reactor for hydrogen enhancement and CO reduction. We have synthesized novel CO{sub 2}-selective membranes with high CO{sub 2} permeabilities and high CO{sub 2}/H{sub 2} and CO{sub 2}/CO selectivities by incorporating amino groups in polymer networks. We have also developed a one-dimensional non-isothermal model for the countercurrent WGS membrane reactor. The modeling results have shown that H{sub 2} enhancement (>99.6% H{sub 2} for the steam reforming of methane and >54% H{sub 2} for the autothermal reforming of gasoline with air on a dry basis) via CO{sub 2} removal and CO reduction to 10 ppm or lower are achievable for synthesis gases. With this model, we have elucidated the effects of system parameters, including CO{sub 2}/H{sub 2} selectivity, CO{sub 2} permeability, sweep/feed flow rate ratio, feed temperature, sweep temperature, feed pressure, catalyst activity, and feed CO concentration, on the membrane reactor performance. Based on the modeling study using the membrane data obtained, we showed the feasibility of achieving H{sub 2} enhancement via CO{sub 2} removal, CO reduction to {le} 10 ppm, and high H{sub 2} recovery. Using the membrane synthesized, we have obtained <10 ppm CO in the H{sub 2} product in WGS membrane reactor experiments. From the experiments, we verified the model developed. In addition, we removed CO{sub 2} from a syngas containing 17% CO{sub 2} to about 30 ppm. The CO{sub 2} removal data agreed well with the model developed. The syngas with about 0.1% CO{sub 2} and 1% CO was processed to convert the carbon oxides to methane via methanation to obtain <5 ppm CO in the H{sub 2} product.

  14. REACTOR

    DOEpatents

    Szilard, L.

    1963-09-10

    A breeder reactor is described, including a mass of fissionable material that is less than critical with respect to unmoderated neutrons and greater than critical with respect to neutrons of average energies substantially greater than thermal, a coolant selected from sodium or sodium--potassium alloys, a control liquid selected from lead or lead--bismuth alloys, and means for varying the quantity of control liquid in the reactor. (AEC)

  15. REACTOR

    DOEpatents

    Christy, R.F.

    1961-07-25

    A means is described for co-relating the essential physical requirements of a fission chain reaction in order that practical, compact, and easily controllable reactors can be built. These objects are obtained by employing a composition of fissionsble isotope and moderator in fluid form in which the amount of fissionsble isotcpe present governs the reaction. The size of the reactor is no longer a critical factor, the new criterion being the concentration of the fissionable isotope.

  16. Detecting pin diversion from pressurized water reactors spent fuel assemblies

    DOEpatents

    Ham, Young S.; Sitaraman, Shivakumar

    2017-01-10

    Detecting diversion of spent fuel from Pressurized Water Reactors (PWR) by determining possible diversion including the steps of providing a detector cluster containing gamma ray and neutron detectors, inserting the detector cluster containing the gamma ray and neutron detectors into the spent fuel assembly through the guide tube holes in the spent fuel assembly, measuring gamma ray and neutron radiation responses of the gamma ray and neutron detectors in the guide tube holes, processing the gamma ray and neutron radiation responses at the guide tube locations by normalizing them to the maximum value among each set of responses and taking the ratio of the gamma ray and neutron responses at the guide tube locations and normalizing the ratios to the maximum value among them and producing three signatures, gamma, neutron, and gamma-neutron ratio, based on these normalized values, and producing an output that consists of these signatures that can indicate possible diversion of the pins from the spent fuel assembly.

  17. Commercial Light Water Reactor Tritium Extraction Facility Geotechnical Summary Report

    SciTech Connect

    Lewis, M R

    2000-01-11

    A geotechnical investigation program has been completed for the Circulating Light Water Reactor - Tritium Extraction Facility (CLWR-TEF) at the Savannah River Site (SRS). The program consisted of reviewing previous geotechnical and geologic data and reports, performing subsurface field exploration, field and laboratory testing and geologic and engineering analyses. The purpose of this investigation was to characterize the subsurface conditions for the CLWR-TEF in terms of subsurface stratigraphy and engineering properties for design and to perform selected engineering analyses. The objectives of the evaluation were to establish site-specific geologic conditions, obtain representative engineering properties of the subsurface and potential fill materials, evaluate the lateral and vertical extent of any soft zones encountered, and perform engineering analyses for slope stability, bearing capacity and settlement, and liquefaction potential. In addition, provide general recommendations for construction and earthwork.

  18. Stability monitor experience in German BWRs (boiling water reactor)

    SciTech Connect

    Goldstein, L. ); Fuge, R. ); Seepolt, R. ); Frank, M. )

    1989-11-01

    A digital stability monitor developed by NIS Ingenieurgesellschaft MBH, operable on a personal computer, is in use at three boiling water reactor (BWR) plants in the Federal Republic of Germany (FRG). The device has received the approval of an FRG licensing authority. It has been in operation for 5 yr. The stability monitor is a measurement device used to accurately determine and confirm the boundaries of the exclusion region on the operating power-flow map, and to permit controlled operation through such regions. This is achieved via neutron noise measurements converted to direct readout of decay ratios. It has satisfied regulatory requirements by defining potential unstable operating regions at Kernkraftwerk Isar in the FRG and the high power density Gundremmingen (KRB Units B and C) BWRs.

  19. Light-water breeder reactor (LWBR Development Program)

    DOEpatents

    Beaudoin, B.R.; Cohen, J.D.; Jones, D.H.; Marier, L.J. Jr.; Raab, H.F.

    1972-06-20

    Described is a light-water-moderated and -cooled nuclear breeder reactor of the seed-blanket type characterized by core modules comprising loosely packed blanket zones enriched with fissile fuel and axial zoning in the seed and blanket regions within each core module. Reactivity control over lifetime is achieved by axial displacement of movable seed zones without the use of poison rods in the embodiment illustrated. The seed is further characterized by a hydrogen-to-uranium-233 atom ratio in the range 10 to 200 and a uranium-233-to-thorium-232 atom ratio ranging from 0.012 to 0.200. The seed occupies from 10 to 35 percent of the core volume in the form of one or more individual islands or annuli. (NSA 26: 55130)

  20. Pressurized water reactor fuel assembly subchannel void fraction measurement

    SciTech Connect

    Akiyama, Yoshiei; Hori, Keiichi; Miyazaki, Keiji; Mishima, Kaichiro; Sugiyama, Shigekazu

    1995-12-01

    The void fraction measurement experiment of pressurized water reactor (PWR) fuel assemblies has been conducted since 1987 under the sponsorship of the Ministry of International Trade and Industry as a Japanese national project. Two types of test sections are used in this experiment. One is a 5 x 5 array rod bundle geometry, and the other is a single-channel geometry simulating one of the subchannels in the rod bundle. Wide gamma-ray beam scanners and narrow gamma-ray beam computed tomography scanners are used to measure the subchannel void fractions under various steady-state and transient conditions. The experimental data are expected to be used to develop a void fraction prediction model relevant to PWR fuel assemblies and also to verify or improve the subchannel analysis method. The first series of experiments was conducted in 1992, and a preliminary evaluation of the data has been performed. The preliminary results of these experiments are described.

  1. Radial nodalization effects on BWR (boiling water reactor) stability calculations

    SciTech Connect

    March-Leuba, J.

    1990-01-01

    Computer simulations have shown that stability calculations in boiling water reactors (BWRs) are very sensitive to a number of input parameters and modeling assumptions. In particular, the number of thermohydraulic regions (i.e., channels) used in the calculation can affect the results of decay ratio calculations by as much as 30%. This paper presents the background theory behind the observed effects of radial nodalization in BWR stability calculations. The theory of how a radial power distribution can be simulated in time or frequency domain codes by using representative'' regions is developed. The approximations involved in this method of solution are reviewed, and some examples of the effect of radial nodalization are presented based on LAPUR code solutions. 2 refs., 4 figs., 2 tabs.

  2. Technology Implementation Plan. Fully Ceramic Microencapsulated Fuel for Commercial Light Water Reactor Application

    SciTech Connect

    Snead, Lance Lewis; Terrani, Kurt A.; Powers, Jeffrey J.; Worrall, Andrew; Robb, Kevin R.; Snead, Mary A.

    2015-04-01

    This report is an overview of the implementation plan for ORNL's fully ceramic microencapsulated (FCM) light water reactor fuel. The fully ceramic microencapsulated fuel consists of tristructural isotropic (TRISO) particles embedded inside a fully dense SiC matrix and is intended for utilization in commercial light water reactor application.

  3. 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

    ... COMMISSION Initial Test Program of Condensate and Feedwater Systems for Light-Water Reactors AGENCY: Nuclear...-1265, ``Initial Test Program of Condensate and Feedwater Systems for Light- Water Reactors.'' DG-1265... plant startup, and power ascension tests for the condensate and feedwater systems in all types of...

  4. 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

    ... COMMISSION Initial Test Program of Condensate and Feedwater Systems for Light-Water Reactors AGENCY: Nuclear... plant startup, and power ascension tests for the condensate and feedwater systems in all types of light..., including condensate storage and supply, for light-water reactors (LWRs) and for startup feedwater...

  5. Final Report on Isotope Ratio Techniques for Light Water Reactors

    SciTech Connect

    Gerlach, David C.; Gesh, Christopher J.; Hurley, David E.; Mitchell, Mark R.; Meriwether, George H.; Reid, Bruce D.

    2009-07-01

    The Isotope Ratio Method (IRM) is a technique for estimating the energy or plutonium production in a fission reactor by measuring isotope ratios in non-fuel reactor components. The isotope ratios in these components can then be directly related to the cumulative energy production with standard reactor modeling methods.

  6. Application of fully ceramic microencapsulated fuels in light water reactors

    SciTech Connect

    Gentry, C.; George, N.; Maldonado, I.; Godfrey, A.; Terrani, K.; Gehin, J.

    2012-07-01

    This study performs a preliminary evaluation of the feasibility of incorporation of Fully Ceramic Microencapsulated (FCM) fuels in light water reactors (LWRs). In particular, pin cell, lattice, and full core analyses are carried out on FCM fuel in a pressurized water reactor (PWR). Using uranium-based fuel and Pu/Np-based fuel in TRistructural isotropic (TRISO) particle form, each fuel design was examined using the SCALE 6.1 analytical suite. In regards to the uranium-based fuel, pin cell calculations were used to determine which fuel material performed best when implemented in the fuel kernel as well as the size of the kernel and surrounding particle layers. The higher fissile material density of uranium mononitride (UN) proved to be favorable, while the parametric studies showed that the FCM particle fuel design with 19.75% enrichment would need roughly 12% additional fissile material in comparison to that of a standard UO{sub 2} rod in order to match the lifetime of an 18-month PWR cycle. As part of the fuel assembly design evaluations, fresh feed lattices were modeled to analyze the within-assembly pin power peaking. Also, a 'color-set' array of assemblies was constructed to evaluate power peaking and power sharing between a once-burned and a fresh feed assembly. In regards to the Pu/Np-based fuel, lattice calculations were performed to determine an optimal lattice design based on reactivity behavior, pin power peaking, and isotopic content. After obtaining a satisfactory lattice design, the feasibility of core designs fully loaded with Pu/Np FCM lattices was demonstrated using the NESTLE three-dimensional core simulator. (authors)

  7. Application of Fully Ceramic Microencapsulated Fuels in Light Water Reactors

    SciTech Connect

    Gentry, Cole A; George, Nathan M; Maldonado, G Ivan; Godfrey, Andrew T; Terrani, Kurt A; Gehin, Jess C

    2012-01-01

    This study aims to perform a preliminary evaluation of the feasibility of incorporation of Fully Ceramic Microencapsulated (FCM) fuels in Light Water Reactors (LWRs). In particular pin cell, lattice, and full core analyses are carried out on FCM fuel in a pressurized water reactor. Using uranium-based fuel and transuranic (TRU) based fuel in TRistructural ISOtropic (TRISO) particle form, each fuel design was examined using the SCALE 6.1 analytical suite. In regards to the uranium-based fuel, pin cell calculations were used to determine which fuel material performed best when implemented in the fuel kernel as well as the size of the kernel and surrounding particle layers. The higher physical density of uranium mononitride (UN) proved to be favorable, while the parametric studies showed that the FCM particle fuel design would need roughly 12% additional fissile material in comparison to that of a standard UO2 rod in order to match the lifetime of an 18-month PWR cycle. As part of the fuel assembly design evaluations, fresh feed lattices were modeled to analyze the within-assembly pin power peaking. Also, a color-set array of assemblies was constructed to evaluate power peaking and power sharing between a once-burned and a fresh feed assembly. In regards to the TRU based fuel, lattice calculations were performed to determine an optimal lattice design based on reactivity behavior, pin power peaking, and isotopic content. After obtaining a satisfactory lattice design, feasibility of core designs fully loaded with TRU FCM lattices was demonstrated using the NESTLE three-dimensional core simulator.

  8. Recent performance experience with US light water reactor self-actuating safety and relief valves

    SciTech Connect

    Hammer, C.G.

    1996-12-01

    Over the past several years, there have been a number of operating reactor events involving performance of primary and secondary safety and relief valves in U.S. Light Water Reactors. There are several different types of safety and relief valves installed for overpressure protection of various safety systems throughout a typical nuclear power plant. The following discussion is limited to those valves in the reactor coolant systems (RCS) and main steam systems of pressurized water reactors (PWR) and in the RCS of boiling water reactors (BWR), all of which are self-actuating having a setpoint controlled by a spring-loaded disk acting against system fluid pressure. The following discussion relates some of the significant recent experience involving operating reactor events or various testing data. Some of the more unusual and interesting operating events or test data involving some of these designs are included, in addition to some involving a number of similar events and those which have generic applicability.

  9. Method and apparatus for separation of heavy and tritiated water

    DOEpatents

    Lee, Myung W.

    2001-01-01

    The present invention is a bi-thermal membrane process for separating and recovering hydrogen isotopes from a fluid containing hydrogen isotopes, such as water and hydrogen gas. The process in accordance with the present invention provides counter-current cold and hot streams of the fluid separated with a thermally insulating and chemically transparent proton exchange membrane (PEM). The two streams exchange hydrogen isotopes through the membrane: the heavier isotopes migrate into the cold stream, while the lighter isotopes migrate into the hot stream. The heavy and light isotopes are continuously withdrawn from the cold and hot streams respectively.

  10. The effects of fire temperatures on water soluble heavy metals.

    NASA Astrophysics Data System (ADS)

    Pereira, P.; Ubeda, X.; Martin, D. A.

    2009-04-01

    Fire ash are majority composed by base cations, however the mineralized organic matter, led also available to transport a higher quantity of heavy metals that potentially could increase a toxicity in soil and water resources. The amount availability of these elements depend on the environment were the fire took place, burning temperature and combusted tree specie. The soil and water contamination from fire ash has been neglected, because the majority of studies are focused on base cations dynamic. Our research, beside contemplate major elements, is focused in to study the behavior of heavy metals released from ash slurries created at several temperatures under laboratory environment, prescribed fires and wildland fires. The results presented in these communication are preliminary and study the presence of Aluminium (Al3+), Manganese (Mn2+), Iron (Fe2+) and Zinc (Zn2+) of ash slurries generated in laboratory environment at several temperatures (150°, 200°, 250°, 300°, 350°, 400°,450°, 500°, 550°C) from Quercus suber, Quercus robur, Pinus pinea and Pinus pinaster and from a low medium temperature prescribed fire in a forest dominated Quercus suber trees. We observed that ash produced at lower and medium temperatures (<300-400°C) released in water higher contents of Al3+ than unburned sample, especially in Quercus species and Mn2+ in Pinus ashes. Fe2+ and Zn2+ showed a reduced concentration in test solution in relation to unburned sample at all temperatures of exposition. In the results obtained from prescribed fire, we identify a higher release of Al3+ and a decrease of the remain elements. The solubilization of these elements are related with pH levels and ash calcite content, because their ability to capture ions in solution. Moreover, the amount and the type of ions released in relation to unburned sample vary in each specie. In this study Al3+ release is related with Quercus species and Mn2+ with Pinus species. Fire ashes can be an environmental problem

  11. Heavy metal removal from waste waters by ion flotation.

    PubMed

    Polat, H; Erdogan, D

    2007-09-05

    Flotation studies were carried out to investigate the removal of heavy metals such as copper (II), zinc (II), chromium (III) and silver (I) from waste waters. Various parameters such as pH, collector and frother concentrations and airflow rate were tested to determine the optimum flotation conditions. Sodium dodecyl sulfate and hexadecyltrimethyl ammonium bromide were used as collectors. Ethanol and methyl isobutyl carbinol (MIBC) were used as frothers. Metal removal reached about 74% under optimum conditions at low pH. At basic pH it became as high as 90%, probably due to the contribution from the flotation of metal precipitates.

  12. New limit on the proton life-time independent on channel from the neutrino experiments with heavy water

    NASA Astrophysics Data System (ADS)

    Tretyak, V. I.; Zdesenko, Yu. G.

    2002-07-01

    Experimental data on the number of neutrons born in the heavy water targets of the large neutrino detectors are used to set the limit on the proton life-time independently on decay mode through the reaction d → n + ?. The best up-to-date limit τ p > 4×10 23 yr with 95% C.L. is derived from the measurements with D 2O target (mass 267 kg) installed near the Bugey reactor. This value can be improved by six orders of magnitude with future data accumulated with the SNO detector containing 1000 t of D 2O.

  13. Heavy Metal Uptakes by Myriophyllum verticillatum from Two Environmental Matrices: The Water and the Sediment.

    PubMed

    Sapci, Zehra; Ustun, E Beyza

    2015-01-01

    To determine the preferred elements of the benthic plant Myriophyllum verticillatum, changes in the element concentrations in the plant were investigated in laboratory condition. The reactor was fed with synthetically contaminated water consisting of 2×10(-6) M of the heavy metals Fe, Cr, Zn, Ni, and Cu for 1060 hours. The elements that were preferentially taken up by the tested plant body were evaluated with respect to translocation factor, bio-concentration factor, and the amounts of partial elements and relative uptakes. Both the changing physical properties of the aqueous solution in the reactor during the experiment and the growth of the plant were tested using a two-sample t-test. The Zn and Cu levels in the combination of the leaves and stems were found to be significantly higher than the levels in the roots at the end of the trial. Based on the partial amount of each element, the affinity of the plant for different elements was found to follow the order of Ca>Fe>Mn. Scanning electron microscope (SEM) analyses of the plant bodies indicated that these elements were located both inside the organs and on the surface of the tissues alone or with microorganisms such as diatoms.

  14. REACTORS

    DOEpatents

    Spitzer, L. Jr.

    1961-10-01

    Thermonuclear reactors, methods, and apparatus are described for controlling and confining high temperature plasma. Main axial confining coils in combination with helical windings provide a rotational transform that avoids the necessity of a figure-eight shaped reactor tube. The helical windings provide a multipolar helical magnetic field transverse to the axis of the main axial confining coils so as to improve the effectiveness of the confining field by counteracting the tendency of the more central lines of force in the stellarator tube to exchange positions with the magnetic lines of force nearer the walls of the tube. (AEC)

  15. Thorium-Based Transmuter Fuels for Light Water Reactors

    SciTech Connect

    J. Stephen Herring; P. E. MacDonald; K. Weaver

    2004-04-01

    A light water reactor (LWR) fuel cycle is proposed where the reactor core mainly consists of standard uranium-dioxide (UO2) fuel rods with typical 235U enrichment, along with thoria-urania (ThO2-UO2) or yttria-stablized zirconia fertile-free fuel rods containing the plutonium and minor actinides typical of 30-yr old UO2 fuel in 1/9 to 1/3 of the positions. The goals of this mono-recycling strategy or "twice through fuel cycle" are to transmute the great majority of the long lived actinides in existing LWRs and to discharge a fuel form that is a very robust waste form and whose isotopic content is very proliferation resistant. The incorporation of plutonium into a ThO2 or yttria-stablized zirconia fertile-free matrix results in the consumption of already-separated plutonium without breeding significant additional 239Pu. The minor actinides (i.e., neptunium, americium, curium, berkelium, californium, etc.) are also included in the ThO2 or fertile-free transmuter fuel rods to further reduce the overall long-term radiotoxicity of the fuel cycle. Our analyses have shown that thorium-based or fertile-free fuels can reduce the amount of 239Pu needing further transmutation or going to a repository by ~90%. Also, thorium-based fuels produce a mixture of plutonium isotopes high in 238Pu. Because of the high decay heat and spontaneous neutron generation of 238Pu, this isotope provides intrinsic proliferation resistance.

  16. Thorium-Based Transmuter Fuels for Light Water Reactors

    SciTech Connect

    Herring, J. Stephen; MacDonald, Philip E.; Weaver, Kevan D.

    2004-07-15

    A light water reactor (LWR) fuel cycle is proposed where the reactor core mainly consists of standard uranium-dioxide (UO{sub 2}) fuel rods with typical {sup 235}U enrichment, along with thoria-urania (ThO{sub 2}-UO{sub 2}) or yttria-stablized zirconia fertile-free fuel rods containing the plutonium and minor actinides typical of 30-yr old UO{sub 2} fuel in 1/9 to 1/3 of the positions. The goals of this mono-recycling strategy or 'twice through fuel cycle' are to transmute the great majority of the long lived actinides in existing LWRs and to discharge a fuel form that is a very robust waste form and whose isotopic content is very proliferation resistant. The incorporation of plutonium into a ThO{sub 2} or yttria-stablized zirconia fertile-free matrix results in the consumption of already-separated plutonium without breeding significant additional {sup 239}Pu. The minor actinides (i.e., neptunium, americium, curium, berkelium, californium, etc.) are also included in the ThO{sub 2} or fertile-free transmuter fuel rods to further reduce the overall long-term radiotoxicity of the fuel cycle. Our analyses have shown that thorium-based or fertile-free fuels can reduce the amount of {sup 239}Pu needing further transmutation or going to a repository by {approx}90%. Also, thorium-based fuels produce a mixture of plutonium isotopes high in {sup 238}Pu. Because of the high decay heat and spontaneous neutron generation of {sup 238}Pu, this isotope provides intrinsic proliferation resistance.

  17. Concentration and speciation of heavy metals during water hyacinth composting.

    PubMed

    Singh, Jiwan; Kalamdhad, Ajay S

    2012-11-01

    The Tessier sequential extraction method was employed to investigate the changes in heavy metals speciation (Zn, Cu, Mn, Fe, Pb, Ni, Cd and Cr) during water hyacinth (Eichhornia crassipes) composting. Results showed that, the contents of total metals concentration were increased during the composting process. The largest proportion of metals was found in the residual fraction which was in more stable form and is consequently considered unavailable for plant uptake. Reducible and oxidizable fractions of Ni, Pb and Cd were not found in all trials during water hyacinth composting. The concentrations of Cu and Cd were very low comparative to the other metals, but the percentage of exchangeable and carbonate fractions were similar as other metals. From this study it can be concluded that the appropriate proportion of cattle manure addition (Trial 4) significantly reduced the mobile and easily available fractions (exchangeable and carbonate fractions) during the composting process. Copyright © 2012 Elsevier Ltd. All rights reserved.

  18. Light-water breeder reactors: preliminary safety and environmental information document. Volume III

    SciTech Connect

    Not Available

    1980-01-01

    Information is presented concerning prebreeder and breeder reactors based on light-water-breeder (LWBR) Type 1 modules; light-water backfit prebreeder supplying advanced breeder; light-water backfit prebreeder/seed-blanket breeder system; and light-water backfit low-gain converter using medium-enrichment uranium, supplying a light-water backfit high-gain converter.

  19. Leaching of heavy metals from water bottle components into the drinking water of rodents.

    PubMed

    Nunamaker, Elizabeth A; Otto, Kevin J; Artwohl, James E; Fortman, Jeffrey D

    2013-01-01

    Providing high-quality, uncontaminated drinking water is an essential component of rodent husbandry. Acidification of drinking water is a common technique to control microbial growth but is not a benign treatment. In addition to its potential biologic effects, acidified water might interact with the water-delivery system, leading to the leaching of heavy metals into the drinking water. The goal of the current study was to evaluate the effects of water acidification and autoclaving on water-bottle assemblies. The individual components of the system (stainless-steel sipper tubes, rubber stoppers, neoprene stoppers, and polysulfone water bottles) were acid-digested and analyzed for cadmium, chromium, copper, iron, lead, magnesium, manganese, selenium, and zinc to quantify the metal composition of each material. In addition the amounts of these metals that leached into tap and acidified water with and without autoclaving were quantified after 1 wk of contact time. On a weight basis, sipper tubes contained the largest quantities of all metals except magnesium and zinc, which were greatest in the neoprene stoppers. Except for cadmium and selenium, all metals had leached into the water after 1 wk, especially under the acidified condition. The quantities of copper, lead, and zinc that leached into the drinking water were the most noteworthy, because the resulting concentrations had the potential to confound animal experiments. On the basis of these findings, we suggest that water-quality monitoring programs include heavy metal analysis at the level of water delivery to animals.

  20. Leaching of Heavy Metals from Water Bottle Components into the Drinking Water of Rodents

    PubMed Central

    Nunamaker, Elizabeth A; Otto, Kevin J; Artwohl, James E; Fortman, Jeffrey D

    2013-01-01

    Providing high-quality, uncontaminated drinking water is an essential component of rodent husbandry. Acidification of drinking water is a common technique to control microbial growth but is not a benign treatment. In addition to its potential biologic effects, acidified water might interact with the water-delivery system, leading to the leaching of heavy metals into the drinking water. The goal of the current study was to evaluate the effects of water acidification and autoclaving on water-bottle assemblies. The individual components of the system (stainless-steel sipper tubes, rubber stoppers, neoprene stoppers, and polysulfone water bottles) were acid-digested and analyzed for cadmium, chromium, copper, iron, lead, magnesium, manganese, selenium, and zinc to quantify the metal composition of each material. In addition the amounts of these metals that leached into tap and acidified water with and without autoclaving were quantified after 1 wk of contact time. On a weight basis, sipper tubes contained the largest quantities of all metals except magnesium and zinc, which were greatest in the neoprene stoppers. Except for cadmium and selenium, all metals had leached into the water after 1 wk, especially under the acidified condition. The quantities of copper, lead, and zinc that leached into the drinking water were the most noteworthy, because the resulting concentrations had the potential to confound animal experiments. On the basis of these findings, we suggest that water-quality monitoring programs include heavy metal analysis at the level of water delivery to animals. PMID:23562029

  1. Simultaneous removal of oil and grease, and heavy metals from artificial bilge water using electro-coagulation/flotation.

    PubMed

    Rincón, Guillermo J; La Motta, Enrique J

    2014-11-01

    US and international regulations pertaining to the control of bilge water discharges from ships have concentrated their attention to the levels of oil and grease rather than to the heavy metal concentrations. The consensus is that any discharge of bilge water (and oily water emulsion within 12 nautical miles from the nearest land cannot exceed 15 parts per million (ppm). Since there is no specific regulation for metal pollutants under the bilge water section, reference standards regulating heavy metal concentrations are taken from the ambient water quality criteria to protect aquatic life. The research herein presented discusses electro-coagulation (EC) as a method to treat bilge water, with a focus on oily emulsions and heavy metals (copper, nickel and zinc) removal efficiency. Experiments were run using a continuous flow reactor, manufactured by Ecolotron, Inc., and a synthetic emulsion as artificial bilge water. The synthetic emulsion contained 5000 mg/L of oil and grease, 5 mg/L of copper, 1.5 mg/L of nickel, and 2.5 mg/l of zinc. The experimental results demonstrate that EC is very efficient in removing oil and grease. For oil and grease removal, the best treatment and cost efficiency was obtained when using a combination of carbon steel and aluminum electrodes, at a detention time less than one minute, a flow rate of 1 L/min and 0.6 A/cm(2) of current density. The final effluent oil and grease concentration, before filtration, was always less than 10 mg/L. For heavy metal removal, the combination of aluminum and carbon steel electrodes, flow rate of 1 L/min, effluent recycling, and 7.5 amps produced 99% zinc removal efficiency. Copper and nickel are harder to remove, and a removal efficiency of 70% was achieved. Copyright © 2014 Elsevier Ltd. All rights reserved.

  2. O^- channels of Dissociative Electron Attachment to water and heavy water molecules

    NASA Astrophysics Data System (ADS)

    Adaniya, Hidehito; Rudek, Benedikt; Osipov, Timur; Lee, Sun; Weber, Thorsten; Hertlein, Marcus; Schoeffler, Markus; Prior, Mike; Belkacem, Ali

    2009-05-01

    A COLTRIM technique is modified to measure the kinetic energy and angular distribution of O^- ions arising from dissociative electron attachment to water and heavy water molecules. A low energy pulsed electron, an effusive water target, a pulsed extraction plate are used in combination with the COLTRIMS spectrometer. The spectrometer carries an electrostatic lens system to compensate the effusiveness of the target. This technique is applied to study the O^- channels in the three Feshbach resonances of water and heavy water anion. The measured kinetic energy release will give the energy partitioning among the fragments, and the means to identify the two-body and three-body breakup channels. The angular distribution of the O^- ions with respect to the electron beam is found to reflect well the breakup dynamics of the H2O^- at the dissociation. The experimental results are compared with the theoretical predictions.

  3. Competitive sorption of heavy metals by water hyacinth roots.

    PubMed

    Zheng, Jia-Chuan; Liu, Hou-Qi; Feng, Hui-Min; Li, Wen-Wei; Lam, Michael Hon-Wah; Lam, Paul Kwan-Sing; Yu, Han-Qing

    2016-12-01

    Heavy metal pollution is a global issue severely constraining aquaculture practices, not only deteriorating the aquatic environment but also threatening the aquaculture production. One promising solution is adopting aquaponics systems where a synergy can be established between aquaculture and aquatic plants for metal sorption, but the interactions of multiple metals in such aquatic plants are poorly understood. In this study, we investigated the absorption behaviors of Cu(II) and Cd(II) in water by water hyacinth roots in both single- and binary-metal systems. Cu(II) and Cd(II) were individually removed by water hyacinth roots at high efficiency, accompanied with release of protons and cations such as Ca(2+) and Mg(2+). However, in a binary-metal arrangement, the Cd(II) sorption was significantly inhibited by Cu(II), and the higher sorption affinity of Cu(II) accounted for its competitive sorption advantage. Ionic exchange was identified as a predominant mechanism of the metal sorption by water hyacinth roots, and the amine and oxygen-containing groups are the main binding sites accounting for metal sorption via chelation or coordination. This study highlights the interactive impacts of different metals during their sorption by water hyacinth roots and elucidates the underlying mechanism of metal competitive sorption, which may provide useful implications for optimization of phytoremediation system and development of more sustainable aquaculture industry. Copyright © 2016 Elsevier Ltd. All rights reserved.

  4. 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 Commission... Systems (RTNSS) for Passive Advanced Light Water Reactors.'' The current SRP does not contain guidance on the proposed RTNSS for Passive Advance Light Water Reactors. DATES: Submit comments by November...

  5. 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

    ... COMMISSION Proposed Revision to Treatment of Non-Safety Systems for Passive Advanced Light Water Reactors... Treatment of Non-Safety Systems (RTNSS) for Passive Advanced Light Water Reactors.'' The NRC seeks public...- Safety Systems (RTNSS) for Passive Advanced Light Water Reactors.'' This area includes a revised...

  6. 10 CFR Appendix A to Part 52 - Design Certification Rule for the U.S. Advanced Boiling Water Reactor

    Code of Federal Regulations, 2010 CFR

    2010-01-01

    ... Water Reactor A Appendix A to Part 52 Energy NUCLEAR REGULATORY COMMISSION (CONTINUED) LICENSES... Rule for the U.S. Advanced Boiling Water Reactor I. Introduction Appendix A constitutes the standard design certification for the U.S. Advanced Boiling Water Reactor (ABWR) design, in accordance with 10 CFR...

  7. 10 CFR Appendix A to Part 52 - Design Certification Rule for the U.S. Advanced Boiling Water Reactor

    Code of Federal Regulations, 2011 CFR

    2011-01-01

    ... Water Reactor A Appendix A to Part 52 Energy NUCLEAR REGULATORY COMMISSION (CONTINUED) LICENSES... Rule for the U.S. Advanced Boiling Water Reactor I. Introduction Appendix A constitutes the standard design certification for the U.S. Advanced Boiling Water Reactor (ABWR) design, in accordance with 10 CFR...

  8. Criticality experiments and analysis of molybdenum reflected cylindrical uranyl fluoride water solution reactors

    NASA Technical Reports Server (NTRS)

    Fieno, D.; Fox, T.; Mueller, R.

    1972-01-01

    Clean criticality data were obtained from molybdenum-reflected cylindrical uranyl-fluoride-water solution reactors. Using ENDF/B molybdenum cross sections, a nine energy group two-dimensional transport calculation of a reflected reactor configuration predicted criticality to within 7 cents of the experimental value. For these reactors, it was necessary to compute the reflector resonance integral by a detailed transport calculation at the core-reflector interface volume in the energy region of the two dominant resonances of natural molybdenum.

  9. Heavy metal removal from copper smelting effluent using electrochemical cylindrical flow reactor.

    PubMed

    Ahmed Basha, C; Bhadrinarayana, N S; Anantharaman, N; Meera Sheriffa Begum, K M

    2008-03-21

    The purpose of this study is mainly to evaluate the performance of the continuous recirculation flow cell at low current density and pH (the pH at which the effluents are available) in removing heavy metals from copper smelting effluent by cathodic reduction. During the electrolysis at different pH, % removal of heavy metals removal, energy consumption and heterogeneous reaction rate constants were investigated at given flow rate and current density on the selected industrial effluent. The overall specific energy consumption at the pH 0.64 was observed to be lowest, which is 10.99kWh/kg of heavy metal removal.

  10. [Development of the heterogeneous photocatalytic reactor in water treatment].

    PubMed

    He, X; Wang, Y

    2001-07-01

    Based on the history and the functions of the heterogenous photocatalytic reactors, three categories were discussed. The emphasis was put on the employment of the reactors designed for the practice in recent years. It was pointed out that the study and the design of the reactors were one of the cores in the process of the application of the photocatalytic oxidation techniques. And the trend of this technology was also predicted.

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

    DOEpatents

    Forsberg, Charles W.

    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.

  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. An evaluation of alternative reactor vessel cutting technologies for the experimental boiling water reactor at Argonne National Laboratory

    SciTech Connect

    Boing, L.E.; Henley, D.R. ); Manion, W.J.; Gordon, J.W. )

    1989-12-01

    Metal cutting techniques that can be used to segment the reactor pressure vessel of the Experimental Boiling Water Reactor (EBWR) at Argonne National Laboratory (ANL) have been evaluated by Nuclear Energy Services. Twelve cutting technologies are described in terms of their ability to perform the required task, their performance characteristics, environmental and radiological impacts, and cost and schedule considerations. Specific recommendations regarding which technology should ultimately be used by ANL are included. The selection of a cutting method was the responsibility of the decommissioning staff at ANL, who included a relative weighting of the parameters described in this document in their evaluation process. 73 refs., 26 figs., 69 tabs.

  14. 77 FR 16098 - In the Matter of All Operating Boiling Water Reactor Licensees With Mark I and Mark II...

    Federal Register 2010, 2011, 2012, 2013, 2014

    2012-03-19

    ... Communications.'' The Director, Office of Nuclear Reactor Regulation may, in writing, relax or rescind any of the... writing to the Director, Office of Nuclear Reactor Regulation, U.S. Nuclear Regulatory Commission... Nuclear Reactor Regulation. Operating Boiling Water Reactor Licenses With Mark I and Mark II...

  15. Effects of diverse water pipe materials on bacterial communities and water quality in the annular reactor.

    PubMed

    Jang, Hyun-Jung; Choi, Young-June; Ka, Jong-Ok

    2011-02-01

    To investigate the effects of pipe materials on biofilm accumulation and water quality, an annular reactor with the sample coupons of four pipe materials (steel, copper, stainless steel, and polyvinyl chloride) was operated under hydraulic conditions similar to a real plumbing system for 15 months. The bacterial concentrations were substantially increased in the steel and copper reactors with progression of corrosion, whereas those in stainless steel (STS) and polyvinyl chloride (PVC) reactors were affected mainly by water temperature. The heterotrophic plate count (HPC) of biofilms was about 100 times higher on steel pipe than other pipes throughout the experiment, with the STS pipe showing the lowest bacterial number at the end of the operation. Analysis of the 16S rDNA sequences of 176 cultivated isolates revealed that 66.5% was Proteobacteria and the others included unclassified bacteria, Actinobacteria, and Bacilli. Regardless of the pipe materials, Sphingomonas was the predominant species in all biofilms. PCR-DGGE analysis showed that steel pipe exhibited the highest bacterial diversity among the metallic pipes, and the DGGE profile of biofilm on PVC showed three additional bands not detected from the profiles of the metallic materials. Environmental scanning electron microscopy showed that corrosion level and biofilm accumulation were the least in the STS coupon. These results suggest that the STS pipe is the best material for plumbing systems in terms of the microbiological aspects of water quality.

  16. Multi-Application Small Light Water Reactor Final Report

    SciTech Connect

    Modro, S.M.; Fisher, J.E.; Weaver, K.D.; Reyes, J.N.; Groome, J.T.; Babka, P.; Carlson, T.M.

    2003-12-01

    The Multi-Application Small Light Water Reactor (MASLWR) project was conducted under the auspices of the Nuclear Energy Research Initiative (NERI) of the U.S. Department of Energy (DOE). The primary project objectives were to develop the conceptual design for a safe and economic small, natural circulation light water reactor, to address the economic and safety attributes of the concept, and to demonstrate the technical feasibility by testing in an integral test facility. This report presents the results of the project. After an initial exploratory and evolutionary process, as documented in the October 2000 report, the project focused on developing a modular reactor design that consists of a self-contained assembly with a reactor vessel, steam generators, and containment. These modular units would be manufactured at a single centralized facility, transported by rail, road, and/or ship, and installed as a series of self-contained units. This approach also allows for staged construction of an NPP and ''pull and replace'' refueling and maintenance during each five-year refueling cycle. Development of the baseline design concept has been sufficiently completed to determine that it complies with the safety requirements and criteria, and satisfies the major goals already noted. The more significant features of the baseline single-unit design concept include: (1) Thermal Power--150 MWt; (2) Net Electrical Output--35 MWe; (3) Steam Generator Type--Vertical, helical tubes; (4) Fuel UO{sub 2}, 8% enriched; (5) Refueling Intervals--5 years; (6) Life-Cycle--60 years. The economic performance was assessed by designing a power plant with an electric generation capacity in the range of current and advanced evolutionary systems. This approach allows for direct comparison of economic performance and forms a basis for further evaluation, economic and technical, of the proposed design and for the design evolution towards a more cost competitive concept. Applications such as

  17. Safe new reactor for radionuclide production

    SciTech Connect

    Gray, P.L.

    1995-02-15

    In late 1995, DOE is schedule to announce a new tritium production unit. Near the end of the last NPR (New Production Reactors) program, work was directed towards eliminating risks in current designs and reducing effects of accidents. In the Heavy Water Reactor Program at Savannah River, the coolant was changed from heavy to light water. An alternative, passively safe concept uses a heavy-water-filled, zircaloy reactor calandria near the bottom of a swimming pool; the calandria is supported on a light-water-coolant inlet plenum and has upflow through assemblies in the calandria tubes. The reactor concept eliminates or reduces significantly most design basis and severe accidents that plague other deigns. The proven, current SRS tritium cycle remains intact; production within the US of medical isotopes such as Mo-99 would also be possible.

  18. Comparative assessment of nuclear fuel cycles. Light-water reactor once-through, classical fast breeder reactor, and symbiotic fast breeder reactor cycles

    SciTech Connect

    Hardie, R.W.; Barrett, R.J.; Freiwald, J.G.

    1980-06-01

    The object of the Alternative Nuclear Fuel Cycle Study is to perform comparative assessments of nuclear power systems. There are two important features of this study. First, this evaluation attempts to encompass the complete, integrated fuel cycle from mining of uranium ore to disposal of waste rather than isolated components. Second, it compares several aspects of each cycle - energy use, economics, technological status, proliferation, public safety, and commercial potential - instead of concentrating on one or two assessment areas. This report presents assessment results for three fuel cycles. These are the light-water reactor once-through cycle, the fast breeder reactor on the classical plutonium cycle, and the fast breeder reactor on a symbiotic cycle using plutonium and /sup 233/U as fissile fuels. The report also contains a description of the methodology used in this assessment. Subsequent reports will present results for additional fuel cycles.

  19. Radiological Control of Water in Reactor Pond of MR Reactor in NRC 'Kurchatov Institute', During Dismantling Work - 13462

    SciTech Connect

    Stepanov, Alexey; Simirsky, Yury; Semin, Ilya; Volkovich, Anatoly; Ivanov, Oleg

    2013-07-01

    The analysis of the activity and radionuclide composition of water from the MR reactor pond for α,β,γ-ray radionuclides was made. To solve this problem we use a wide range of laboratory equipment: gamma spectrometric complex, beta spectrometric complex, vacuum alpha spectrometer, and spectrometric complex with liquid scintillator. The water from MR reactor pond contains: Cs-137 (2,6*10{sup 2} Bq/g), Co-60(1,8 Bq/g), Sr-90 (1,0*10{sup 2} Bq/g), H-3 (7,0*10{sup 3} Bq/g), and components of nuclear fuel (U-232,U-234,U-235,U-236,U-238). Therefore the cleaning water from radioactivity waste occurs to be quite a complicated radiochemical task. (authors)

  20. Design of GA thermochemical water-splitting process for the Mirror Advanced Reactor System

    SciTech Connect

    Brown, L.C.

    1983-04-01

    GA interfaced the sulfur-iodine thermochemical water-splitting cycle to the Mirror Advanced Reactor System (MARS). The results of this effort follow as one section and part of a second section to be included in the MARS final report. This section describes the process and its interface to the reactor. The capital and operating costs for the hydrogen plant are described.

  1. Grouping of light water reactors for evaluation of decay heat removal capability

    SciTech Connect

    Karol, R.; Fresco, A.; Perkins, K.R.

    1984-06-01

    This grouping report provides a compilation of decay heat removal systems (DHRS) data for operating commercial light water reactors. The reactors have been divided into 12 groups based on similarity of the DHRS and related systems as part of the NRC Task Action Plan on Shutdown Decay Heat Removal Requirements.

  2. GENERIC, COMPONENT FAILURE DATA BASE FOR LIGHT WATER AND LIQUID SODIUM REACTOR PRAs

    SciTech Connect

    S. A. Eide; S. V. Chmielewski; T. D. Swantz

    1990-02-01

    A comprehensive generic component failure data base has been developed for light water and liquid sodium reactor probabilistic risk assessments (PRAs) . The Nuclear Computerized Library for Assessing Reactor Reliability (NUCLARR) and the Centralized Reliability Data Organization (CREDO) data bases were used to generate component failure rates . Using this approach, most of the failure rates are based on actual plant data rather than existing estimates .

  3. Water Reactor Safety Research Division quarterly progress report, January 1-March 31, 1980

    SciTech Connect

    Romano, A.J.

    1980-06-01

    The Water Reactor Safety Research Programs Quarterly Report describes current activities and technical progress in the programs at Brookhaven National Laboratory sponsored by the USNRC Division of Reactor Safety Research. The projects reported each quarter are the following: LWR Thermal Hydraulic Development, Advanced Code Evaluation, TRAC Code Assessment, and Stress Corrosion Cracking of PWR Steam Generator Tubing.

  4. Water Reactor Safety Research Division. Quarterly progress report, April 1-June 30, 1980

    SciTech Connect

    Abuaf, N.; Levine, M.M.; Saha, P.; van Rooyen, D.

    1980-08-01

    The Water Reactor Safety Research Programs quarterly report describes current activities and technical progress in the programs at Brookhaven National Laboratory sponsored by the USNRC Division of Reactor Safety Research. The projects reported each quarter are the following: LWR Thermal Hydraulic Development, Advanced Code Evlauation, TRAC Code Assessment, and Stress Corrosion Cracking of PWR Steam Generator Tubing.

  5. Pollution Status of Pakistan: A Retrospective Review on Heavy Metal Contamination of Water, Soil, and Vegetables

    PubMed Central

    Arshad, Jahanzaib; Iqbal, Farhat; Sajjad, Ashif; Mehmood, Zahid

    2014-01-01

    Trace heavy metals, such as arsenic, cadmium, lead, chromium, nickel, and mercury, are important environmental pollutants, particularly in areas with high anthropogenic pressure. In addition to these metals, copper, manganese, iron, and zinc are also important trace micronutrients. The presence of trace heavy metals in the atmosphere, soil, and water can cause serious problems to all organisms, and the ubiquitous bioavailability of these heavy metal can result in bioaccumulation in the food chain which especially can be highly dangerous to human health. This study reviews the heavy metal contamination in several areas of Pakistan over the past few years, particularly to assess the heavy metal contamination in water (ground water, surface water, and waste water), soil, sediments, particulate matter, and vegetables. The listed contaminations affect the drinking water quality, ecological environment, and food chain. Moreover, the toxicity induced by contaminated water, soil, and vegetables poses serious threat to human health. PMID:25276818

  6. Pollution status of Pakistan: a retrospective review on heavy metal contamination of water, soil, and vegetables.

    PubMed

    Waseem, Amir; Arshad, Jahanzaib; Iqbal, Farhat; Sajjad, Ashif; Mehmood, Zahid; Murtaza, Ghulam

    2014-01-01

    Trace heavy metals, such as arsenic, cadmium, lead, chromium, nickel, and mercury, are important environmental pollutants, particularly in areas with high anthropogenic pressure. In addition to these metals, copper, manganese, iron, and zinc are also important trace micronutrients. The presence of trace heavy metals in the atmosphere, soil, and water can cause serious problems to all organisms, and the ubiquitous bioavailability of these heavy metal can result in bioaccumulation in the food chain which especially can be highly dangerous to human health. This study reviews the heavy metal contamination in several areas of Pakistan over the past few years, particularly to assess the heavy metal contamination in water (ground water, surface water, and waste water), soil, sediments, particulate matter, and vegetables. The listed contaminations affect the drinking water quality, ecological environment, and food chain. Moreover, the toxicity induced by contaminated water, soil, and vegetables poses serious threat to human health.

  7. Heavy metals in drinking water: Occurrences, implications, and future needs in developing countries.

    PubMed

    Chowdhury, Shakhawat; Mazumder, M A Jafar; Al-Attas, Omar; Husain, Tahir

    2016-11-01

    Heavy metals in drinking water pose a threat to human health. Populations are exposed to heavy metals primarily through water consumption, but few heavy metals can bioaccumulate in the human body (e.g., in lipids and the gastrointestinal system) and may induce cancer and other risks. To date, few thousand publications have reported various aspects of heavy metals in drinking water, including the types and quantities of metals in drinking water, their sources, factors affecting their concentrations at exposure points, human exposure, potential risks, and their removal from drinking water. Many developing countries are faced with the challenge of reducing human exposure to heavy metals, mainly due to their limited economic capacities to use advanced technologies for heavy metal removal. This paper aims to review the state of research on heavy metals in drinking water in developing countries; understand their types and variability, sources, exposure, possible health effects, and removal; and analyze the factors contributing to heavy metals in drinking water. This study identifies the current challenges in developing countries, and future research needs to reduce the levels of heavy metals in drinking water. Copyright © 2016 Elsevier B.V. All rights reserved.

  8. Impact of inflow transport approximation on light water reactor analysis

    NASA Astrophysics Data System (ADS)

    Choi, Sooyoung; Smith, Kord; Lee, Hyun Chul; Lee, Deokjung

    2015-10-01

    The impact of the inflow transport approximation on light water reactor analysis is investigated, and it is verified that the inflow transport approximation significantly improves the accuracy of the transport and transport/diffusion solutions. A methodology for an inflow transport approximation is implemented in order to generate an accurate transport cross section. The inflow transport approximation is compared to the conventional methods, which are the consistent-PN and the outflow transport approximations. The three transport approximations are implemented in the lattice physics code STREAM, and verification is performed for various verification problems in order to investigate their effects and accuracy. From the verification, it is noted that the consistent-PN and the outflow transport approximations cause significant error in calculating the eigenvalue and the power distribution. The inflow transport approximation shows very accurate and precise results for the verification problems. The inflow transport approximation shows significant improvements not only for the high leakage problem but also for practical large core problem analyses.

  9. Depletion optimization of lumped burnable poisons in pressurized water reactors

    SciTech Connect

    Kodah, Z.H.

    1982-01-01

    Techniques were developed to construct a set of basic poison depletion curves which deplete in a monotonical manner. These curves were combined to match a required optimized depletion profile by utilizing either linear or non-linear programming methods. Three computer codes, LEOPARD, XSDRN, and EXTERMINATOR-2 were used in the analyses. A depletion routine was developed and incorporated into the XSDRN code to allow the depletion of fuel, fission products, and burnable poisons. The Three Mile Island Unit-1 reactor core was used in this work as a typical PWR core. Two fundamental burnable poison rod designs were studied. They are a solid cylindrical poison rod and an annular cylindrical poison rod with water filling the central region.These two designs have either a uniform mixture of burnable poisons or lumped spheroids of burnable poisons in the poison region. Boron and gadolinium are the two burnable poisons which were investigated in this project. Thermal self-shielding factor calculations for solid and annular poison rods were conducted. Also expressions for overall thermal self-shielding factors for one or more than one size group of poison spheroids inside solid and annular poison rods were derived and studied. Poison spheroids deplete at a slower rate than the poison mixture because each spheroid exhibits some self-shielding effects of its own. The larger the spheroid, the higher the self-shielding effects due to the increase in poison concentration.

  10. Improved pressurized water reactor radial reflector modeling in nodal analysis

    SciTech Connect

    Mueller, E.Z. )

    1991-10-01

    A one-dimensional method based on a combination of the nodal equivalence theory and response matrix homogenization methods was previously described for determining environment-insensitive equivalent few-group diffusion theory parameters for homogenized radial reflector nodes of a pressurized water reactor. This reflector model, called the NGET-RM model, yields equivalent nodal parameters that do not account for the two-dimensional structure of the baffle at core corners; this can lead to significant errors in computed two-dimensional core power distributions. A semi-empirical correction procedure is proposed for reducing the two-dimensional effects associated with this particular one-dimensional reflector model. Numerical two-group experiments are performed for a given reflector configuration (and soluble boron concentration) to determine optimal values of the two empirical factors defined by this model. In this paper it is shown that the resultant factors are rather insensitive to core configuration or core conditions and that their application yields improved two-group NGET-RM reflector parameters with which accurate nodal power distributions can be obtained. The results are also compared with those obtained with another one-dimensional environment-insensitive model that has an extra degree of freedom utilized here to reduce two-dimensional effects. Some practical aspects related to the application of the proposed correction procedure are briefly discussed.

  11. Multi-cycle boiling water reactor fuel cycle optimization

    SciTech Connect

    Ottinger, K.; Maldonado, G.I.

    2013-07-01

    In this work a new computer code, BWROPT (Boiling Water Reactor Optimization), is presented. BWROPT uses the Parallel Simulated Annealing (PSA) algorithm to solve the out-of-core optimization problem coupled with an in-core optimization that determines the optimum fuel loading pattern. However it uses a Haling power profile for the depletion instead of optimizing the operating strategy. The result of this optimization is the optimum new fuel inventory and the core loading pattern for the first cycle considered in the optimization. Several changes were made to the optimization algorithm with respect to other nuclear fuel cycle optimization codes that use PSA. Instead of using constant sampling probabilities for the solution perturbation types throughout the optimization as is usually done in PSA optimizations the sampling probabilities are varied to get a better solution and/or decrease runtime. The new fuel types available for use can be sorted into an array based on any number of parameters so that each parameter can be incremented or decremented, which allows for more precise fuel type selection compared to random sampling. Also, the results are sorted by the new fuel inventory of the first cycle for ease of comparing alternative solutions. (authors)

  12. Installation of the Light-Water Breeder Reactor at the Shippingport Atomic Power Station (LWBR Development Program)

    SciTech Connect

    Massimino, R.J.; Williams, D.A.

    1983-05-01

    This report summarizes the refueling operations performed to install a Light Water Breeder Reactor (LWBR) core into the existing pressurized water reactor vessel at the Shippingport Atomic Power Station. Detailed descriptions of the major installation operations (e.g., primary system preconditioning, fuel installation, pressure boundary seal welding) are included as appendices to this report; these operations are of technical interest to any reactor servicing operation, whether the reactor is a breeder or a conventional light water non-breeder core.

  13. General features of direct-cycle, supercritical-pressure, light-water-cooled reactors

    SciTech Connect

    Oka, Y.; Koshizuka, S.

    1996-07-01

    The concept of direct-cycle, supercritical-pressure, light-water-cooled reactors is developed. Breeding is possible in the tight lattice core. The power output can be maximized in the fast converter reactor. The gross thermal efficiency of the high temperature reactor adopting Inconel as fuel cladding is expected to be 44.8%. The plant system is similar to the supercritical-fossil-fired power plant which adopts once-through type coolant circulation system. The volume and height of the containment are approximately half of the BWR. The basic safety principles follows those of LWRs. The reactor will solve the economic problems of LWR and LMFBR.

  14. [Effect of reclaimed water irrigation on soil properties and vertical distribution of heavy metal].

    PubMed

    Zhao, Zong-Ming; Chen, Wei-Ping; Jiao, Wen-Tao; Wang, Mei-E

    2012-12-01

    Utilization of reclaimed water is one of the important methods to alleviate water shortage. The effect of reclaimed water irrigation on soil is always a concern. To understand the effect of long time reclaimed water irrigation on soil, typical farmland irrigated with reused water was selected. Soil properties and heavy metal concentration of soil and water samples were analyzed to identify the effect of the irrigation on heavy metal vertical distribution and organic matter content, total carbon, total nitrogen and pH value in soil. The results show that heavy metal contents of irrigation water used in Liangshuihe farmland are 2.5 to 10.5 times higher than that of Beiyechang farmland, and reclaimed water irrigation could cause changes of soil properties that soil organic matter content, total carbon, total nitrogen were increased and pH values were reduced. Based on the field investigation results, the soil nutrient conditions benefit from irrigate reclaimed water, however, the accumulation of heavy metal in soil could raise the risk. As a source of soil heavy metal, reclaimed water irrigation could make differences on the accumulation and mobility of soil heavy metal. Also the distribution and mobility of soil heavy metal are influenced by soil organic matter content and there are more heavy metal were taken up by plants or transferred to the deeper area in Liangshuihe farmland.

  15. 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.

  16. Models and Stability Analysis of Boiling Water Reactors

    SciTech Connect

    John Dorning

    2002-04-15

    We have studied the nuclear-coupled thermal-hydraulic stability of boiling water reactors (BWRs) using a model that includes: space-time modal neutron kinetics based on spatial w-modes; single- and two-phase flow in parallel boiling channels; fuel rod heat conduction dynamics; and a simple model of the recirculation loop. The BR model is represented by a set of time-dependent nonlinear ordinary differential equations, and is studied as a dynamical system using the modern bifurcation theory and nonlinear dynamical systems analysis. We first determine the stability boundary (SB) - or Hopf bifurcation set- in the most relevant parameter plane, the inlet-subcooling-number/external-pressure-drop plane, for a fixed control rod induced external reactivity equal to the 100% rod line value; then we transform the SB to the practical power-flow map used by BWR operating engineers and regulatory agencies. Using this SB, we show that the normal operating point at 100% power is very stable, that stability of points on the 100% rod line decreases as the flow rate is reduced, and that operating points in the low-flow/high-power region are least stable. We also determine the SB that results when the modal kinetics is replaced by simple point reactor kinetics, and we thereby show that the first harmonic mode does not have a significant effect on the SB. However, we later show that it nevertheless has a significant effect on stability because it affects the basin of attraction of stable operating points. Using numerical simulations we show that, in the important low-flow/high-power region, the Hopf bifurcation that occurs as the SB is crossed is subcritical; hence, growing oscillations can result following small finite perturbations of stable steady-states on the 100% rod line at points in the low-flow/high-power region. Numerical simulations are also performed to calculate the decay ratios (DRs) and frequencies of oscillations for various points on the 100% rod line. It is

  17. An integrated approach for the verification of fresh mixed oxide fuel (MOX) assemblies at light water reactor MOX recycle reactors

    SciTech Connect

    Menlove, Howard O; Lee, Sang - Yoon

    2009-01-01

    This paper presents an integrated approach for the verification of mixed oxide (MOX) fuel assemblies prior to their being loaded into the reactor. There is a coupling of the verification approach that starts at the fuel fabrication plant and stops with the transfer of the assemblies into the thermal reactor. The key measurement points are at the output of the fuel fabrication plant, the receipt at the reactor site, and the storage in the water pool as fresh fuel. The IAEA currently has the capability to measure the MOX fuel assemblies at the output of the fuel fabrication plants using a passive neutron coincidence counting systems of the passive neutron collar (PNCL) type. Also. at the MOX reactor pool, the underwater coincidence counter (UWCC) has been developed to measure the MOX assemblies in the water. The UWCC measurement requires that the fuel assembly be lifted about two meters up in the storage rack to avoid interference from the fuel that is stored in the rack. This paper presents a new method to verify the MOX fuel assemblies that are in the storage rack without the necessity of moving the fuel. The detector system is called the Underwater MOX Verification System (UMVS). The integration and relationship of the three measurements systems is described.

  18. Genetic risk assessment of acid waste water containing heavy metals.

    PubMed

    Miadoková, E; Dúhová, V; Vlcková, V; Sládková, L; Sucha, V; Vlcek, D

    1999-10-01

    The mutagenic/cancerogenic potential of acid-mine water from the Slovak mining area Rudnany containing a high load of toxic metals was evaluated after its application to three model test organisms (bacteria Salmonella typhimurium, yeast Saccharomyces cerevisiae and plant Vicia sativa L.). The results obtained from the modified preincubation Ames assay proved that 1000-fold diluted waste water exhibited mutagenic effect in three (TA97, TA98, TA102) of four bacterial strains. In the test on yeast the toxicity and genotoxicity increased as a function of the concentration. At the highest concentration used (0.06%) the frequency of revertants increased 6 times and convertants increased 4.5 times above the control level. In the simultaneous phytotoxicity and clastogenicity assay, concentration dependent toxicity and statistically significant clastogenicity was proved. We can conclude that heavy metals might be responsible for the genotoxic/cancerogenic potential of the test water. However, we do not entirely exclude the possibility that its genotoxicity might be promoted by its high acidity.

  19. Nondestructive assay of spent boiling water reactor fuel by active neutron interrogation

    SciTech Connect

    Blakeman, E.D.; Ricker, C.W.; Ragan, G.L.; Difilippo, F.C.; Slaughter, G.G.

    1981-01-01

    Spent boiling water reactor (BWR) fuel from Dresden I was assayed for total fissile mass, using the active neutron interrogation method. The nondestructive assay (NDA) system used has four Sb-Be sources for interrogation of the fuels; the induced fission neutrons from the fuel are counted by four lead-shielded methane-filled proportional counters biased above the energy of the source neutrons. Spent fuel rods containing 9 kg of heavy metal were chopped into 5-cm segments and loaded into three 1-liter cans. The three cans were assayed in seven combinations of one, two, or three cans, enabling an evaluation of the precision and accuracy of the NDA system for different amounts of fissile material. The fissile mass in each combination was determined by comparing the induced-fission-neutron counts with the counts obtained from a known standard comprising chopped segments of unirradiated Dresden fuel. These masses were compared to the masses determined by chemical analyses of the spent fuel. The results from the nondestructive assays agreed with results from the chemical analyses to within 2 to 3%. Similar agreement was obtained when two combinations of canned spent fuel were used as standards for the nondesctuctive assays. The assay of BWR spent fuel served as a test of the NDA system which was developed at the Oak Ridge National Laboratory for the assay of spent liquid metal fast breeder reactor (LMFBR) fuel subassemblies at the heat-end of a reprocessing plant. Results of previous experiments and calculations reported earlier using simulated LMFBR fuel subassemblies indicated that the NDA system can measure the fissile masses of spent fuel subassemblies to within an accuracy of 3%. Results of the assays of spent BWR fuel reported herein support this conclusion.

  20. Survey of light-water-reactor designs to be offered in the United States

    SciTech Connect

    Spiewak, I.

    1986-03-01

    ORNL has conducted a Nuclear Power Options Viability Study for the Department of Energy. That study is primarily concerned with new technology which could be developed for initial operation in the 2000 to 2010 time frame. Such technology would have to compete not only with coal options but with incrementally improved commercial light-water-reactors. This survey reported here was undertaken to gain an understanding of the nuclear commercial technology likely to be offered in the late 1980s and perhaps beyond. The three US vendors actively marketing NSSSs are each developing a product for the future which they expect to be more reliable, more maintainable, more economical, and safer than the present plants. These are all essentially 3800-MW(t) designs, although all are studying smaller plants. They apparently will be off offered as standard prelicensed designs with much larger scope than earlier NSSS offerings, with the possibility of firm prices. Westinghouse with Mitsubishi Heavy Industries is developing a completely new design (APWR) to be built initially in Japan, hopefully for operation by the mid-1990s. Westinghouse is making a strong effort to have the APWR licensed in the US as a standard plant. Combustion Engineering (C-E) is evaluating potential improvements to the System-80 standard design (CESSAR) that has already received final design approval by the NRC. General Electric (GE), with Hitachi and Toshiba, is developing a new design (ABWR) that incorporates advanced features which have been proven by the worldwide BWR suppliers. The ABWR is to be built initially in Japan, but the design could be adapted to the United States. Westinghouse, C-E, and GE have done some conceptual evaluation of reactors in the 600-MW(e) class. The Westinghouse concept is a two-loop plant intended for factory assembly in a shipyard and delivery to a site by barge. The GE concept is a modification of the ABWR with some additional passive safety features. 16 figs.

  1. Heavy metals in vegetables and respective soils irrigated by canal, municipal waste and tube well waters.

    PubMed

    Ismail, Amir; Riaz, Muhammad; Akhtar, Saeed; Ismail, Tariq; Amir, Mamoona; Zafar-ul-Hye, Muhammad

    2014-01-01

    Heavy metal contamination in the food chain is of serious concern due to the potential risks involved. The results of this study revealed the presence of maximum concentration of heavy metals in the canal followed by sewerage and tube well water. Similarly, the vegetables and respective soils irrigated with canal water were found to have higher heavy metal contamination followed by sewerage- and tube-well-watered samples. However, the heavy metal content of vegetables under study was below the limits as set by FAO/WHO, except for lead in canal-water-irrigated spinach (0.59 mg kg(-1)), radish pods (0.44 mg kg(-1)) and bitter gourd (0.33 mg kg(-1)). Estimated daily intakes of heavy metals by the consumption of selected vegetables were found to be well below the maximum limits. However, a complete estimation of daily intake requires the inclusion of other dietary and non-dietary exposure sources of heavy metals.

  2. Pressurized hydrogenotrophic denitrification reactor for small water systems.

    PubMed

    Epsztein, Razi; Beliavski, Michael; Tarre, Sheldon; Green, Michal

    2017-03-15

    The implementation of hydrogenotrophic denitrification is limited due to safety concerns, poor H2 utilization and low solubility of H2 gas with the resulting low transfer rate. The current paper presents the main research work conducted on a pressurized hydrogenotrophic reactor for denitrification that was recently developed. The reactor is based on a new concept suggesting that a gas-liquid equilibrium is achieved in the closed headspace of denitrifying reactor, further produced N2 gas is carried out by the effluent and gas purging is not required. The feasibility of the proposed reactor was shown for two effluent concentrations of 10 and 1 mg NO3(-)-N/L. Hydrogen gas utilization efficiencies of 92.8% and 96.9% were measured for the two effluent concentrations, respectively. Reactor modeling predicted high denitrification rates above 4 g NO3(-)-N/(Lreactor·d) at reasonable operational conditions. Hydrogen utilization efficiency was improved up to almost 100% by combining the pressurized reactor with a following open-to-atmosphere polishing unit. Also, the potential of the reactor to remove ClO4(-) was shown.

  3. Coupled reactor physics and coolant dynamics of heavy liquid metal coolant systems.

    SciTech Connect

    Cahalan, J. E.; Dunn, F. E.; Taiwo, T. A.

    1999-07-15

    Cooling of advanced nuclear designs with heavy liquid metals such as lead or lead-bismuth eutectic offers the potential for plant simplifications and higher operating efficiencies compared to previously considered liquid metal coolants such as sodium or NaK. Such applications would however also introduce additional safety concerns and design challenges, therefore necessitating a verifiable computational tool for transient design-basis analysis of heavy liquid metal coolant (HLMC) systems. This capability would enable analysts to compare operational and safety characteristics of design alternatives, and to evaluate relative performance advantages with a consistent, deterministic measure.

  4. Irradiation effects in low-alloy reactor pressure vessel steels (Heavy-Section Steel Technology Program Series 4 and 5)

    SciTech Connect

    Berggren, R.G.; McGowan, J.J.; Menke, B.H.; Nanstad, R.K.; Thoms, K.R.

    1984-01-01

    Multiple testing is done at two laboratories of typical nuclear pressure vessel materials (both irradiated and unirradiated) and statistical analyses of the test results. Multiple tests are conducted at each of several test temperatures for each material, standard deviations are determined, and results from the two laboratories are compared. The Fourth Heavy-Section Steel Technology (HSST) Irradiation Series, almost completed, was aimed at elastic-plastic and fully plastic fracture toughness of low-copper weldments (current practice welds). A typical nuclear pressure vessel plate steel was included for statistical purposes. The Fifth HSST Irradiation Series, now in progress, is aimed at determining the shape of the K/sub IR/ curve after significant radiation-induced shift of the transition temperatures. This series includes irradiated test specimens of thicknesses up to 100 mm and weldment compositions typical of early nuclear power reactor pressure vessel welds.

  5. Prospects for development of an innovative water-cooled nuclear reactor for supercritical parameters of coolant

    NASA Astrophysics Data System (ADS)

    Kalyakin, S. G.; Kirillov, P. L.; Baranaev, Yu. D.; Glebov, A. P.; Bogoslovskaya, G. P.; Nikitenko, M. P.; Makhin, V. M.; Churkin, A. N.

    2014-08-01

    The state of nuclear power engineering as of February 1, 2014 and the accomplished elaborations of a supercritical-pressure water-cooled reactor are briefly reviewed, and the prospects of this new project are discussed based on this review. The new project rests on the experience gained from the development and operation of stationary water-cooled reactor plants, including VVERs, PWRs, BWRs, and RBMKs (their combined service life totals more than 15 000 reactor-years), and long-term experience gained around the world with operation of thermal power plants the turbines of which are driven by steam with supercritical and ultrasupercritical parameters. The advantages of such reactor are pointed out together with the scientific-technical problems that need to be solved during further development of such installations. The knowledge gained for the last decade makes it possible to refine the concept and to commence the work on designing an experimental small-capacity reactor.

  6. Light Water Reactor Sustainability Program: Analysis of Pressurized Water Reactor Station Blackout Caused by External Flooding Using the RISMC Toolkit

    SciTech Connect

    Smith, Curtis; Mandelli, Diego; Prescott, Steven; Alfonsi, Andrea; Rabiti, Cristian; Cogliati, Joshua; Kinoshita, Robert

    2014-08-01

    The existing fleet of nuclear power plants is in the process of extending its lifetime and increasing the power generated from these plants via power uprates. In order to evaluate the impact of these factors on the safety of the plant, the Risk Informed Safety Margin Characterization (RISMC) project aims to provide insight to decision makers through a series of simulations of the plant dynamics for different initial conditions (e.g., probabilistic analysis and uncertainty quantification). This report focuses, in particular, on the application of a RISMC detailed demonstration case study for an emergent issue using the RAVEN and RELAP-7 tools. This case study looks at the impact of a couple of challenges to a hypothetical pressurized water reactor, including: (1) a power uprate, (2) a potential loss of off-site power followed by the possible loss of all diesel generators (i.e., a station black-out event), (3) and earthquake induces station-blackout, and (4) a potential earthquake induced tsunami flood. The analysis is performed by using a set of codes: a thermal-hydraulic code (RELAP-7), a flooding simulation tool (NEUTRINO) and a stochastic analysis tool (RAVEN) – these are currently under development at the Idaho National Laboratory.

  7. The National Shipbuilding Research Program, Heavy Metal Adsorbents for Storm Water Pollution Prevention

    DTIC Science & Technology

    1997-12-01

    Heavy Metal Adsorbents for Storm Water Pollution Prevention U.S. DEPARTMENT OF THE NAVY CARDEROCK DIVISION, NAVAL SURFACE WARFARE CENTER in...National Shipbuilding Research Program, Heavy Metal Adsorbents for Storm Water Pollution Prevention 5a. CONTRACT NUMBER 5b. GRANT NUMBER 5c. PROGRAM...States Navy. ANY POSSIBLE IMPLIED WARRANTIES OF MERCHANTABILITY AND/OR FITNESS FOR PURPOSE ARE SPECIFICALLY DISCLAIMED. FINAL REPORT HEAVY METAL ADSORBENTS

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

    DOEpatents

    McDermott, Daniel J.; Schrader, Kenneth J.; Schulz, Terry L.

    1994-01-01

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

  9. 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.

  10. Light-water-reactor safety research program. Quarterly progress report, January-March 1980

    SciTech Connect

    Massey, W.E.; Kyger, J.A.

    1980-08-01

    This progress report summarizes the Argonne National Laboratory work performed during January, February, and March 1980 on water-reactor-safety problems. The research and development area covered is Transient Fuel Response and Fission-Product Release.

  11. NUCLEAR REACTOR

    DOEpatents

    Treshow, M.

    1958-08-19

    A neuclear reactor is described of the heterogeneous type and employing replaceable tubular fuel elements and heavy water as a coolant and moderator. A pluraltty of fuel tubesa having their axes parallel, extend through a tank type pressure vessel which contatns the liquid moderator. The fuel elements are disposed within the fuel tubes in the reaetive portion of the pressure vessel during normal operation and the fuel tubes have removable plug members at each end to permit charging and discharging of the fuel elements. The fuel elements are cylindrical strands of jacketed fissionable material having helical exterior ribs. A bundle of fuel elements are held within each fuel tube with their longitudinal axes parallel, the ribs serving to space them apart along their lengths. Coolant liquid is circulated through the fuel tubes between the spaced fuel elements. Suitable control rod and monitoring means are provided for controlling the reactor.

  12. 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.

  13. Feasibility Study of Supercritical Light Water Cooled Reactors for Electric Power Production

    SciTech Connect

    Philip MacDonald; Jacopo Buongiorno; James Sterbentz; Cliff Davis; Robert Witt; Gary Was; J. McKinley; S. Teysseyre; Luca Oriani; Vefa Kucukboyaci; Lawrence Conway; N. Jonsson: Bin Liu

    2005-02-13

    The supercritical water reactor (SCWR) has been the object of interest throughout the nuclear Generation IV community because of its high potential: a simple, direct cycle, compact configuration; elimination of many traditional LWR components, operation at coolant temperatures much higher than traditional LWRs and thus high thermal efficiency. It could be said that the SWR was viewed as the water counterpart to the high temperature gas reactor.

  14. Light Water Reactor-Pressure Vessel Surveillance project computer system

    SciTech Connect

    Merriman, S.H.

    1980-10-01

    A dedicated process control computer has been implemented for regulating the metallurgical Pressure Vessel Wall Benchmark Facility (PSF) at the Oak Ridge Research Reactor. The purpose of the PSF is to provide reliable standards and methods by which to judge the radiation damage to reactor pressure vessel specimens. Benchmark data gathered from the PSF will be used to improve and standardize procedures for assessing the remaining safe operating lifetime of aging reactors. The computer system controls the pressure vessel specimen environment in the presence of gamma heating so that in-vessel conditions are simulated. Instrumented irradiation capsules, in which the specimens are housed, contain temperature sensors and electrical heaters. The computer system regulates the amount of power delivered to the electrical heaters based on the temperature distribution within the capsules. Time-temperature profiles are recorded along with reactor conditions for later correlation with specimen metallurgical changes.

  15. Synergism of Pseudomonas aeruginosa and Fe0 for treatment of heavy metal contaminated effluents using small scale laboratory reactor.

    PubMed

    Singh, Rajesh; Bishnoi, Narsi R; Kirrolia, Anita; Kumar, Rajender

    2013-01-01

    In this study Pseudomonas aeruginosa a metal tolerant strain was not only applied for heavy metal removal but also to the solublization performance of the precipitated metal ions during effluent treatment. The synergistic effect of the isolate and Fe(0) enhanced the metal removal potential to 72.97% and 87.63% for Cr(VI) and cadmium, respectively. The decrease in cadmium ion removal to 43.65% (aeration+stirring reactors), 21.33% (aerated reactors), and 18.95% (without aerated+without stirring) with an increase in incubation period not only indicate the presence of soluble less toxic complexes, but also help in exploration of the balancing potential for valuable metal recovery. A relatively best fit and significant values of the correlation coefficient 0.912, 0.959, and 0.9314 for mixed effluent (Paint Industry effluent+CETP Wazirpur, effluent), CETP, Wazirpur, and control effluents, respectively, indicating first-order formulation and provide a reasonable description of COD kinetic data. Copyright © 2012 Elsevier Ltd. All rights reserved.

  16. Reactor physics and safety aspects of various design options of a Russian light water reactor with rock-like fuels

    NASA Astrophysics Data System (ADS)

    Bondarenko, A. V.; Komissarov, O. V.; Kozmenkov, Ya. K.; Matveev, Yu. V.; Orekhov, Yu. I.; Pivovarov, V. A.; Sharapov, V. N.

    2003-06-01

    This paper presents results of analytical studies on weapons grade plutonium incineration in VVER (640) medium size light water reactors using a special composition of rock-like fuel (ROX-fuel) to assure spent fuel long-term storage without its reprocessing. The main goal is to achieve high degree of plutonium incineration in once-through cycle. In this paper we considered two fuel compositions. In both compositions weapons grade plutonium is used as fissile material. Spinel (MgAl 2O 4) is used as the 'preserving' material assuring safe storage of the spent fuel. Besides an inert matrix, the option of rock-like fuel with thorium dioxide was studied. One of principal problems in the realization of the proposed approach is the substantial change of properties of the light water reactor core when passing to the use of the ROX-fuel, in particular: (i) due to the absence of 238U the Doppler effect playing a crucial role in reactor's self-regulation and limiting the consequences of reactivity accidents, decreases significantly, (ii) no fuel breeding on one hand, and the quest to attain the maximum plutonium burnup on the other hand, would result in a drastical change of the fuel assembly power during the lifetime and, as a consequence, the rise in irregularity of the power density of fuel assemblies, (iii) both the control rods worth and dissolved boron worth decrease in view of neutron spectrum hardening brought on by the larger absorption cross-section of plutonium as compared to uranium, (iv) βeff is markedly reduced. All these distinctive features are potentially detrimental to the reactor nuclear safety. The principal objective of this work is that to identify a variant of the fuel composition and the reactor layout, which would permit neutralize the negative effect of the above-mentioned distinctive features.

  17. Passive gamma analysis of the boiling-water-reactor assemblies

    SciTech Connect

    Vo, D.; Favalli, A.; Grogan, B.; Jansson, P.; Liljenfeldt, H.; Mozin, V.; Schwalbach, P.; Sjöland, A.; Tobin, S.; Trellue, H.; Vaccaro, S.

    2016-09-01

    This research focused on the analysis of a set of stationary passive gamma measurements taken on the spent nuclear fuel assemblies from a boiling water reactor (BWR) using pulse height analysis data acquisition. The measurements were performed on 25 different BWR assemblies in 2014 at Sweden’s Central Interim Storage Facility for Spent Nuclear Fuel (Clab). This study was performed as part of the Next Generation of Safeguards Initiative–Spent Fuel project to research the application of nondestructive assay (NDA) to spent fuel assemblies. The NGSI–SF team is working to achieve the following technical goals more easily and efficiently than in the past using nondestructive assay (NDA) measurements of spent fuel assemblies: (1) verify the initial enrichment, burnup, and cooling time of facility declaration; (2) detect the diversion or replacement of pins, (3) estimate the plutonium mass, (4) estimate the decay heat, and (5) determine the reactivity of spent fuel assemblies. The final objective of this project is to quantify the capability of several integrated NDA instruments to meet the aforementioned goals using the combined signatures of neutrons, gamma rays, and heat. This report presents a selection of the measured data and summarizes an analysis of the results. Specifically, trends in the count rates measured for spectral lines from the following isotopes were analyzed as a function of the declared burnup and cooling time: 137Cs, 154Eu, 134Cs, and to a lesser extent, 106Ru and 144Ce. From these measured count rates, predictive algorithms were developed to enable the estimation of the burnup and cooling time. Furthermore, these algorithms were benchmarked on a set of assemblies not included in the standard assemblies set used by this research team.

  18. Passive gamma analysis of the boiling-water-reactor assemblies

    DOE PAGES

    Vo, D.; Favalli, A.; Grogan, B.; ...

    2016-09-01

    This research focused on the analysis of a set of stationary passive gamma measurements taken on the spent nuclear fuel assemblies from a boiling water reactor (BWR) using pulse height analysis data acquisition. The measurements were performed on 25 different BWR assemblies in 2014 at Sweden’s Central Interim Storage Facility for Spent Nuclear Fuel (Clab). This study was performed as part of the Next Generation of Safeguards Initiative–Spent Fuel project to research the application of nondestructive assay (NDA) to spent fuel assemblies. The NGSI–SF team is working to achieve the following technical goals more easily and efficiently than in themore » past using nondestructive assay (NDA) measurements of spent fuel assemblies: (1) verify the initial enrichment, burnup, and cooling time of facility declaration; (2) detect the diversion or replacement of pins, (3) estimate the plutonium mass, (4) estimate the decay heat, and (5) determine the reactivity of spent fuel assemblies. The final objective of this project is to quantify the capability of several integrated NDA instruments to meet the aforementioned goals using the combined signatures of neutrons, gamma rays, and heat. This report presents a selection of the measured data and summarizes an analysis of the results. Specifically, trends in the count rates measured for spectral lines from the following isotopes were analyzed as a function of the declared burnup and cooling time: 137Cs, 154Eu, 134Cs, and to a lesser extent, 106Ru and 144Ce. From these measured count rates, predictive algorithms were developed to enable the estimation of the burnup and cooling time. Furthermore, these algorithms were benchmarked on a set of assemblies not included in the standard assemblies set used by this research team.« less

  19. Suboptimal projective control of a pressurized water reactor

    SciTech Connect

    Saif, M. )

    1989-12-01

    The time- and oxide field-dependencies of interface trap (N{sub it}) formation in MOSFETs have been studied following pulsed ionizing radiation. Results are compared with the two-stage model for Nit formation involving slow drift of radiation-induced H{sup +} ions in the SiO{sub 2}. Detailed data on the gate oxide field dependence during each individual stage are presented and discussed. A model is developed for the production of H{sup +} throughout the oxide. Calculations based on this model correctly predict the complete time dependent N{sub it} formation is at a maximum near zero first stage gate bias. This unexpected behavior apparently arises from the oxide field dependence of the H{sup +} production during the first stage. A suboptimal output feedback approach for control of the pressurized water reactor (PWR) in the H. B. Robinson nuclear power plant is presented. Optimal state feedback linear quadratic regulator (LQR) theory with pole placement capability is extended to obtain a suboptimal projective controller for such cases where the entire state vector is inaccessible for measurement and feedback purposes. The appealing feature of the proposed approach is that it is possible to select the weighting matrices in the quadratic cost functional such that the resulting control law would nearly minimize the cost, and at the same time can assign a subspectrum of the closed-loop system to preassigned desired locations. Additionally, the design algorithm is computationally attractive, since regardless of the dimension of the PWR model the approach mainly involves low-order matrix computations.

  20. Passive gamma analysis of the boiling-water-reactor assemblies

    SciTech Connect

    Vo, D.; Favalli, A.; Grogan, B.; Jansson, P.; Liljenfeldt, H.; Mozin, V.; Schwalbach, P.; Sjöland, A.; Tobin, S.; Trellue, H.; Vaccaro, S.

    2016-09-01

    This research focused on the analysis of a set of stationary passive gamma measurements taken on the spent nuclear fuel assemblies from a boiling water reactor (BWR) using pulse height analysis data acquisition. The measurements were performed on 25 different BWR assemblies in 2014 at Sweden’s Central Interim Storage Facility for Spent Nuclear Fuel (Clab). This study was performed as part of the Next Generation of Safeguards Initiative–Spent Fuel project to research the application of nondestructive assay (NDA) to spent fuel assemblies. The NGSI–SF team is working to achieve the following technical goals more easily and efficiently than in the past using nondestructive assay (NDA) measurements of spent fuel assemblies: (1) verify the initial enrichment, burnup, and cooling time of facility declaration; (2) detect the diversion or replacement of pins, (3) estimate the plutonium mass, (4) estimate the decay heat, and (5) determine the reactivity of spent fuel assemblies. The final objective of this project is to quantify the capability of several integrated NDA instruments to meet the aforementioned goals using the combined signatures of neutrons, gamma rays, and heat. This report presents a selection of the measured data and summarizes an analysis of the results. Specifically, trends in the count rates measured for spectral lines from the following isotopes were analyzed as a function of the declared burnup and cooling time: 137Cs, 154Eu, 134Cs, and to a lesser extent, 106Ru and 144Ce. From these measured count rates, predictive algorithms were developed to enable the estimation of the burnup and cooling time. Furthermore, these algorithms were benchmarked on a set of assemblies not included in the standard assemblies set used by this research team.

  1. Passive gamma analysis of the boiling-water-reactor assemblies

    NASA Astrophysics Data System (ADS)

    Vo, D.; Favalli, A.; Grogan, B.; Jansson, P.; Liljenfeldt, H.; Mozin, V.; Schwalbach, P.; Sjöland, A.; Tobin, S.; Trellue, H.; Vaccaro, S.

    2016-09-01

    This research focused on the analysis of a set of stationary passive gamma measurements taken on the spent nuclear fuel assemblies from a boiling water reactor (BWR) using pulse height analysis data acquisition. The measurements were performed on 25 different BWR assemblies in 2014 at Sweden's Central Interim Storage Facility for Spent Nuclear Fuel (Clab). This study was performed as part of the Next Generation of Safeguards Initiative-Spent Fuel project to research the application of nondestructive assay (NDA) to spent fuel assemblies. The NGSI-SF team is working to achieve the following technical goals more easily and efficiently than in the past using nondestructive assay (NDA) measurements of spent fuel assemblies: (1) verify the initial enrichment, burnup, and cooling time of facility declaration; (2) detect the diversion or replacement of pins, (3) estimate the plutonium mass, (4) estimate the decay heat, and (5) determine the reactivity of spent fuel assemblies. The final objective of this project is to quantify the capability of several integrated NDA instruments to meet the aforementioned goals using the combined signatures of neutrons, gamma rays, and heat. This report presents a selection of the measured data and summarizes an analysis of the results. Specifically, trends in the count rates measured for spectral lines from the following isotopes were analyzed as a function of the declared burnup and cooling time: 137Cs, 154Eu, 134Cs, and to a lesser extent, 106Ru and 144Ce. From these measured count rates, predictive algorithms were developed to enable the estimation of the burnup and cooling time. Furthermore, these algorithms were benchmarked on a set of assemblies not included in the standard assemblies set used by this research team.

  2. Systems design of direct-cycle supercritical-water-cooled fast reactors

    SciTech Connect

    Oka, Yoshiaki; Koshizuka, Seiichi; Jevremovic, Tatjana; Okano, Yashushi

    1995-01-01

    The system design of a direct-cycle supercritical-water-cooled fast reactor is presented. The supercritical water does not exhibit a change of phase. the recirculation system, steam separator, and dryer of a boiling water reactor (BWR) are unnecessary. Roughly speaking, the reactor pressure vessel and control rods are similar to those of a pressurized water reactor, the containment and emergency core cooling system are similar to a BWR, and the balance of plant is similar to a supercritical-pressure fossil-fired power plant (FPP). the electric power of the fast converter is 1,508 MW(electric). The number of coolant loops is only two because of the high coolant enthalpy. Containment volume is much reduced. The thermal efficiency is improved 24% over a BWR. The coolant void reactivity is negative by placing thin zirconium-hydride layers between seeds and blankets. The power costs would be much reduced compared with those of a light water reactor (LWR) and a liquid-metal fast breeder reactor. The concept is based on the huge amount of experience with the water coolant technology of LWRs and FPPs. The oxidation of stainless steel cladding is avoided by adopting a much lower coolant temperature than that of the FPP.

  3. Depletion analysis of mixed-oxide fuel pins in light water reactors and the Advanced Test Reactor

    SciTech Connect

    Chang, G.S.; Ryskamp, J.M.

    2000-03-01

    An experiment containing weapons-grade mixed-oxide (WG-MOX) fuel has been designed and is being irradiated in the Advanced Test Reactor (ATR) at the Idaho National Engineering and Environmental Laboratory (INEEL). The ability to accurately predict fuel pin performance is an essential requirement for the MOX fuel test assembly design. Detailed radial fission power and temperature profile effects and fission gas release in the fuel pin are a function of the fuel pin's temperature, fission power, and fission product ad actinide concentration profiles. In addition, the burnup-dependent profile analyses in irradiated fuel pins is important for fuel performance analysis to support the potential licensing of the MOX fuel made from WG-plutonium and depleted uranium for use in US reactors. The MCNP Coupling With ORIGEN2 burnup calculation code (MCWO) can analyze the detailed burnup profiles of WG-MOX and reactor-grade mixed-oxide (RG-MOX) fuel pins. The validated code MCWO can provide the best-estimate neutronic characteristics of fuel burnup performance analysis. Applying this capability with a new minicell method allows calculation of detailed nuclide concentration and power distributions within the MOX pins as a function of burnup. This methodology was applied to MOX fuel in a commercial pressurized water reactor and in an experiment currently being irradiated in the ATR. The prediction of nuclide concentration profiles and power distributions in irradiated MOX plus via this new methodology can provide insights into MOX fuel performance.

  4. Roadmap for Nondestructive Evaluation of Reactor Pressure Vessel Research and Development by the Light Water Reactor Sustainability Program

    SciTech Connect

    Smith, Cyrus M; Nanstad, Randy K; Clayton, Dwight A; Matlack, Katie; Ramuhalli, Pradeep; Light, Glenn

    2012-09-01

    The Department of Energy s (DOE) Light Water Reactor Sustainability (LWRS) Program is a five year effort which works to develop the fundamental scientific basis to understand, predict, and measure changes in materials and systems, structure, and components as they age in environments associated with continued long-term operations of existing commercial nuclear power reactors. This year, the Materials Aging and Degradation (MAaD) Pathway of this program has placed emphasis on emerging Non-Destructive Evaluation (NDE) methods which support these objectives. DOE funded Research and Development (R&D) on emerging NDE techniques to support commercial nuclear reactor sustainability is expected to begin next year. This summer, the MAaD Pathway invited subject matter experts to participate in a series of workshops which developed the basis for the research plan of these DOE R&D NDE activities. This document presents the results of one of these workshops which are the DOE LWRS NDE R&D Roadmap for Reactor Pressure Vessels (RPV). These workshops made a substantial effort to coordinate the DOE NDE R&D with that already underway or planned by the Electric Power Research Institute (EPRI) and the Nuclear Regulatory Commission (NRC) through their representation at these workshops.

  5. Modeling of an annular photocatalytic reactor for water purification: oxidation of pesticides.

    PubMed

    Puma, Gianluca L I; Khor, Jen Nee; Brucato, Alberto

    2004-07-01

    Photocatalytic oxidation (PCO) over titanium dioxide (TiO2) is a "green" sustainable process for the treatment and purification of water and wastewater. However, the application of PCO for wastewater treatment on an industrial scale is currently hindered by a lack of simple mathematical models that can be readily applied to reactor design. Current models are either too simplistic or too rigorous to be useful in photocatalytic reactor design, scale-up, and optimization. In this paper a simple mathematical model is presented for slurry, annular, photocatalytic reactors that still retains the essential elements of a rigorous approach while providing simple solutions. The model extends the applicability of the thin-film model of photocatalytic reactors previously presented to include the case of geometrically thick photoreactors (i.e., those reactors in which the thickness of the annular zone is significant as compared to the outer radius of the reactor). The model uses a novel six-flux absorption-scattering model to represent the radiation field in the reaction space, which assumes that scattered photons follow the route of the six directions of the Cartesian coordinates. The model was successfully validated with experimental results from the photocatalytic oxidation of the pesticide isoproturon in an experimental reactor. The mathematical model presented may be used as a tool for the design, scale-up, and optimization of annular photocatalytic reactors for water treatment and purification.

  6. Compartment in vertical flow reactor for ferruginous mine water

    NASA Astrophysics Data System (ADS)

    Hur, Won; Cheong, Young-Wook; Yim, Gil-Jae; Ji, Sang-Woo; Hong, Ji-Hye

    2014-05-01

    Mine effluents contain varying concentrations of ferrous ion along with other metal ions. Fe(II) that quickly oxidizes to form precipitates in the presence of oxygen under net alkaline or neutral conditions. Thus, passive treatment methods are designed for the mine water to reside in an open containment area so as to allow simultaneous oxidation and precipitation of Fe(II), such as in a lagoon or an oxidation pond. A vertical flow reactor (VFR) was also suggested to remediate ferruginous mine drainage passing down through an accreting bed of ochre. However, VFR has a limited operation time until the system begins to overflow. It was also demonstrated that two-compartment VFR has a longer operation time than single compartment VFR of same size. In this study, a mathematical model was developed as a part of efforts to explore the operation of VFR, showing dynamic changes in head differences, ochre depth and Fe(II)/Fe(III) concentration in the effluent flow. The analysis shows that Fe(II) oxidation and ochre formation should be balanced with permeability of ochre bed to maximize VFR operation time and minimize residual Fe(II) in the effluent. The model demonstrates that two compartment VFR can have a longer operation time than a single-compartment VFR and that an optimum compartment ratio exists that maximize VFR operation time. Accelerated Fe(II) oxidation significantly affects the optimum ratio of compartment area and reduced residual Fe(II) in the effluent. VFR operation time can be significantly prolonged by increasing the rate of ochre formation not by accelerated Fe(II) oxidation. Taken together, ochre forms largely in the first compartment while overflowed mine water with reduced iron contents is efficiently filtered in the second compartment. These results provide us a better understanding of VFR operation and optimum design criteria for maximum operation time in a two-compartment VFR. Rapid ochre accretion in the first compartment maintains constant hydraulic

  7. Validation of NESTLE against static reactor benchmark problems

    SciTech Connect

    Mosteller, R.D.

    1996-02-01

    The NESTLE advanced modal code was developed at North Carolina State University with support from Los Alamos National Laboratory and Idaho National Engineering Laboratory. It recently has been benchmarked successfully against measured data from pressurized water reactors (PWRs). However, NESTLE`s geometric capabilities are very flexible, and it can be applied to a variety of other types of reactors. This study presents comparisons of NESTLE results with those from other codes for static benchmark problems for PWRs, boiling water reactors (BWRs), high-temperature gas-cooled reactors (HTGRs) and CANDU heavy- water reactors (HWRs).

  8. Design Parameters for a Natural Uranium UO{sub 3} or U{sub 3}O{sub 8} Fueled Nuclear Reactor

    SciTech Connect

    Hopper, C.M.

    2002-11-15

    A recent Oak Ridge National Laboratory report provided preliminary analyses to propose alternative design parameters for a nuclear reactor that could be fueled with natural UO{sub 3} or U{sub 3}O{sub 8} and moderated with either heavy water or reactor-grade graphite. This report provides more specific reactor design and operating parameters for a heavy water-moderated reactor only. The basic assumptions and analytical approach are discussed together with the results of the analysis.

  9. Catalytic dewaxing of light and heavy oils in dual parallel reactors

    SciTech Connect

    Chester, A.W.; Garwood, W.E.; Vartuli, J.C.

    1986-08-12

    An integrated process is described for catalytically dewaxing a relatively light petroleum chargestock, characterized by a 50% boiling point of less than about 850/sup 0/F and a kinematic viscosity at 100/sup 0/C of less than about 9 centistokes, and a relatively heavy petroleum chargestock, characterized by a 50% boiling point of greater than about 850/sup 0/F and a kinematic velocity at 100/sup 0/C of greater than about 9 centistokes.

  10. Radiolysis of the coolant in the VK-50 boiling water reactor

    SciTech Connect

    Zabelin, A.I.; Shmelev, V.E.

    1986-10-01

    Radiolysis of the coolant proceeds at a higher rate in a boiling water reactor as compared to a water-moderated, water-cooled reactor. The radiolytic gases (hydrogen and oxygen) exiting the reactor together with steam can form a potentially explosive mixture. Special interest attaches to the results obtained under the codnitions of prolonged operation of the VK-50 reactor. Tests of various water-chemistry conditions which were performed in the experimental reactor showed their critical influence on the rate of progress of radiolytic processes. The entire period of operation of the reactor may be arbitrarily divided into three stages, each of which is characterized by its own peculiar conditions of water chemistry and range of thermal power. From stage to stage, there is a noticeable improvement in the coolant quality which to a limited extent is reflected in the exit of radiolytic gases with the steam. The concentration of radiolytic gases increases with decreased power and with an increased content of corrosion products and other contaminants in the coolant.

  11. Phytoremediation: role of terrestrial plants and aquatic macrophytes in the remediation of radionuclides and heavy metal contaminated soil and water.

    PubMed

    Sharma, Sunita; Singh, Bikram; Manchanda, V K

    2015-01-01

    Nuclear power reactors are operating in 31 countries around the world. Along with reactor operations, activities like mining, fuel fabrication, fuel reprocessing and military operations are the major contributors to the nuclear waste. The presence of a large number of fission products along with multiple oxidation state long-lived radionuclides such as neptunium ((237)Np), plutonium ((239)Pu), americium ((241/243)Am) and curium ((245)Cm) make the waste streams a potential radiological threat to the environment. Commonly high concentrations of cesium ((137)Cs) and strontium ((90)Sr) are found in a nuclear waste. These radionuclides are capable enough to produce potential health threat due to their long half-lives and effortless translocation into the human body. Besides the radionuclides, heavy metal contamination is also a serious issue. Heavy metals occur naturally in the earth crust and in low concentration, are also essential for the metabolism of living beings. Bioaccumulation of these heavy metals causes hazardous effects. These pollutants enter the human body directly via contaminated drinking water or through the food chain. This issue has drawn the attention of scientists throughout the world to device eco-friendly treatments to remediate the soil and water resources. Various physical and chemical treatments are being applied to clean the waste, but these techniques are quite expensive, complicated and comprise various side effects. One of the promising techniques, which has been pursued vigorously to overcome these demerits, is phytoremediation. The process is very effective, eco-friendly, easy and affordable. This technique utilizes the plants and its associated microbes to decontaminate the low and moderately contaminated sites efficiently. Many plant species are successfully used for remediation of contaminated soil and water systems. Remediation of these systems turns into a serious problem due to various anthropogenic activities that have

  12. Study of Pu consumption in advanced light water reactors: Evaluation of GE advanced boiling water reactor plants - compilation of Phase 1B task reports

    SciTech Connect

    1993-09-15

    This report contains an extensive evaluation of GE advanced boiling water reactor plants prepared for United State Department of Energy. The general areas covered in this report are: core and system performance; fuel cycle; infrastructure and deployment; and safety and environmental approval.

  13. Heavy metals content in water, water hyacinth and sediments of Lalbagh tank, Bangalore (India).

    PubMed

    Lokeshwari, H; Chandrappa, G T

    2006-07-01

    The present study was undertaken for assessing the level of heavy metals such as iron, zinc, copper, nickel, chromium, lead and cadmium in water, water hyacinth and sediment samples of Lalbagh tank, Bangalore. Metals were detected using flame atomic absorption spectrophotometry. The results revealed that by and large all metals are present in all the samples, except cadmium in the sediment samples. The average concentrations of iron in water and sediment samples, and lead in water hyacinth were found exceeding the limits of Indian Standards. In general, the concentrations of iron and zinc were found more, followed by lead, chromium, copper, nickel and cadmium cancentration was low. Bioaccumulation factor for water hyacinth was found maximum for chromium. Geoaccumulation index results revealed that there is moderate input of copper and lead from anthropogenic sources to the tank basin.

  14. Assessment of sustainable vermiconversion of water hyacinth at different reactor efficiencies employing Eudrilus eugeniae kinberg.

    PubMed

    Gajalakshmi, S; Ramasamy, E V; Abbasi, S A

    2001-11-01

    The viability of vermireactors fed with different proportions of water hyacinth (WH) and cowdung (CD) was assessed over six-month trials. All reactors performed sustainably with a steadily rising vermicast output, worm zoomass, and number of offspring. There was no mortality in any of the reactors. A change in the WH:CD ratios from 4:1 to 6:1 had no discernable impact on the reactor performances. Attempts were also made to improve the efficiency of the reactors in terms of vermicast production per unit time and per unit digester volume. These attempts led to the 'high-rate' vermireactor in which 5.6 times greater vermicast was produced per litre of digester volume per day than in the 'low-rate' reactors. The high-rate vermireactors also performed sustainably, with steady vermicast output, animal growth, and reproduction.

  15. Heavy metal bioleaching and sludge stabilization in a single-stage reactor using indigenous acidophilic heterotrophs.

    PubMed

    Mehrotra, Akanksha; Sreekrishnan, T R

    2017-01-10

    Simultaneous sludge digestion and metal leaching (SSDML) have been reported at mesophilic temperature. It is generally perceived that while sludge stabilization is effected by heterotrophs at neutral pH, metal bioleaching is done by acidophilic autotrophs. However, little information is available on the microbial communities involved in the process. This study carried out SSDML in a single-stage reactor using sludge indigenous microorganisms and looked at the bacterial communities responsible for the process. Volatile suspended solids were reduced by more than 40%. The concentration of zinc, copper, chromium, cadmium and nickel decreased by more than 45% in the dry sludge. Acidophilic species of Alicyclobacillus genus were the dominant heterotrophs. A few heterotrophic bacteria were detected which can oxidize iron (Alicyclobacillus ferrooxydans, Alicyclobacillus ferripilum and Ferrimicrobium acidiphilum). Acidithiobacillus ferrooxidans (autotroph) was responsible for the oxidation of both iron and sulfur which lead to a change in the pH from neutral to acidic. The presence of acidophilic heterotrophs, which can oxidize either iron or sulfur, enhanced the efficiency of SSDML process with respect to sludge stabilization and metal leaching. This study shows that it is possible to carry out the SSDML in a single-stage reactor with indigenous microorganisms.

  16. Spur decay kinetics of the solvated electron in heavy water radiolysis.

    SciTech Connect

    Bartels, D. M.; Gosztola, D.; Jonah, C. D.; Chemistry

    2001-08-30

    Spur decay kinetics of the hydrated electron following picosecond pulse radiolysis of heavy water have been measured using a time-correlated absorption spectroscopy (TCAS) technique. The TCAS data collected for the first 40 ns of the decay was matched up with single-shot transient digitizer data out to microsecond time scales. The decay shape in heavy water looks exactly like the decay in light water except in the first 10 ns. The 'time zero' solvated electron yield in heavy water radiolysis must be approximately 7% larger than in light water, to match the best available scavenger product measurements. We propose an explanation in terms of the larger distances traveled by electrons in heavy water prior to localization. The implication is that presolvated H{sub 2}O{sup +} 'holes' are very efficient scavengers for the presolvated conduction band electrons.

  17. Light Water Reactor Sustainability Program. Digital Architecture Requirements

    SciTech Connect

    Thomas, Kenneth; Oxstrand, Johanna

    2015-03-01

    The Digital Architecture effort is a part of the Department of Energy (DOE) sponsored Light-Water Reactor Sustainability (LWRS) Program conducted at Idaho National Laboratory (INL). The LWRS program is performed in close collaboration with industry research and development (R&D) programs that provides the technical foundations for licensing and managing the long-term, safe, and economical operation of current nuclear power plants (NPPs). One of the primary missions of the LWRS program is to help the U.S. nuclear industry adopt new technologies and engineering solutions that facilitate the continued safe operation of the plants and extension of the current operating licenses. Therefore, a major objective of the LWRS program is the development of a seamless digital environment for plant operations and support by integrating information from plant systems with plant processes for nuclear workers through an array of interconnected technologies. In order to get the most benefits of the advanced technology suggested by the different research activities in the LWRS program, the nuclear utilities need a digital architecture in place to support the technology. A digital architecture can be defined as a collection of information technology (IT) capabilities needed to support and integrate a wide-spectrum of real-time digital capabilities for nuclear power plant performance improvements. It is not hard to imagine that many processes within the plant can be largely improved from both a system and human performance perspective by utilizing a plant wide (or near plant wide) wireless network. For example, a plant wide wireless network allows for real time plant status information to easily be accessed in the control room, field workers’ computer-based procedures can be updated based on the real time plant status, and status on ongoing procedures can be incorporated into smart schedules in the outage command center to allow for more accurate planning of critical tasks. The goal

  18. Electromyogram as a measure of heavy metal toxicity in fresh water and salt water mussels

    SciTech Connect

    Kidder, G.W. III |; McCoy, A.A. |

    1996-02-01

    The response of bivalves to heavy metals and other toxins has usually been determined by observing valve position. Since mussels close their valves to avoid noxious stimuli, experimental delivery of chemicals ins uncertain. To obtain constant results plastic spacers can be employed to hold the valves apart. This obviates valve position as an index of response and some other method is required. Electromyography of intact mussels is one such index, giving a simple, effective, and quantitative measurement of activity. Experiments are reported in this article on the effects of added mercury on salt water and fresh water species.

  19. Selected bibliography on heavy water, tritiated water and hydrogen isotopes (1981-1992)

    NASA Astrophysics Data System (ADS)

    Gopalakrishnan, V. T.; Sutawane, U. B.; Rathi, B. N.

    A selected bibliography on heavy water, tritiated water and hydrogen isotopes is presented. This bibliography covers the period 1981-1992 and is in continuation to Division's earlier report BARC-1192 (1983). The sources of information for this compilation are Chemical Abstracts, INIS Atom Index and also some scattered search through journals and reports available in our library. No claim is made towards exhaustiveness of this bibliography even though sincere attempts have been made for a wide coverage. The bibliography is arranged under the headings: (1) production, purification, recovery, reprocessing and storage; (2) isotope exchange; (3) isotope analysis; (4) properties; and (5) miscellaneous. Total number of references in the bibliography are 1762.

  20. Comparison of actinide production in traveling wave and pressurized water reactors

    SciTech Connect

    Osborne, A.G.; Smith, T.A.; Deinert, M.R.

    2013-07-01

    The geopolitical problems associated with civilian nuclear energy production arise in part from the accumulation of transuranics in spent nuclear fuel. A traveling wave reactor is a type of breed-burn reactor that could, if feasible, reduce the overall production of transuranics. In one possible configuration, a cylinder of natural or depleted uranium would be subjected to a fast neutron flux at one end. The neutrons would transmute the uranium, producing plutonium and higher actinides. Under the right conditions, the reactor could become critical, at which point a self-stabilizing fission wave would form and propagate down the length of the reactor cylinder. The neutrons from the fission wave would burn the fissile nuclides and transmute uranium ahead of the wave to produce additional fuel. Fission waves in uranium are driven largely by the production and fission of {sup 239}Pu. Simulations have shown that the fuel burnup can reach values greater than 400 MWd/kgIHM, before fission products poison the reaction. In this work we compare the production of plutonium and minor actinides produced in a fission wave to that of a UOX fueled light water reactor, both on an energy normalized basis. The nuclide concentrations in the spent traveling wave reactor fuel are computed using a one-group diffusion model and are verified using Monte Carlo simulations. In the case of the pressurized water reactor, a multi-group collision probability model is used to generate the nuclide quantities. We find that the traveling wave reactor produces about 0.187 g/MWd/kgIHM of transuranics compared to 0.413 g/MWd/kgIHM for a pressurized water reactor running fuel enriched to 4.95 % and burned to 50 MWd/kgIHM. (authors)

  1. Random Mutagenesis by Error-Prone Polymerase Chain Reaction Using a Heavy Water Solvent.

    PubMed

    Minamoto, Toshifumi

    2017-01-01

    Heavy water is a form of water that contains a heavier isotope of hydrogen ((2)H, also known as deuterium, D) or oxygen ((17)O or (18)O). When using heavy water as a solvent, error-prone polymerase chain reaction (epPCR) can induce random mutations independent of the polymerase used or the composition of the PCR reaction mixture. This relatively new method can easily be combined with the existing epPCR methods to increase the rate of mutations.

  2. FLUID MODERATED REACTOR

    DOEpatents

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

    1957-10-22

    A reactor which utilizes fissionable fuel elements in rod form immersed in a moderator or heavy water and a means of circulating the heavy water so that it may also function as a coolant to remove the heat generated by the fission of the fuel are described. In this design, the clad fuel elements are held in vertical tubes immersed in heavy water in a tank. The water is circulated in a closed system by entering near the tops of the tubes, passing downward through the tubes over the fuel elements and out into the tank, where it is drawn off at the bottom, passed through heat exchangers to give up its heat and then returned to the tops of the tubes for recirculation.

  3. Fuzzy logic control of steam generator water level in pressurized water reactors

    SciTech Connect

    Kuan, C.C.; Lin, C.; Hsu, C.C. . Dept. of Nuclear Engineering)

    1992-10-01

    In this paper a fuzzy logic controller is applied to control the steam generator water level in a pressurized water reactor. The method does not require a detailed mathematical mode of the object to be controlled. The design is based on a set of linguistic rules that were adopted from the human operator's experience. After off-line fuzzy computation, the controller is a lookup table, and thus, real-time control is achieved. Shrink-and-swell phenomena are considered in the linguistic rules, and the simulation results show that their effect is dramatically reduced. The performance of the control system can also be improved by changing the input and output scaling factors, which is convenient for on-line tuning.

  4. Overview of the Consortium for the Advanced Simulation of Light Water Reactors (CASL)

    NASA Astrophysics Data System (ADS)

    Kulesza, Joel A.; Franceschini, Fausto; Evans, Thomas M.; Gehin, Jess C.

    2016-02-01

    The Consortium for Advanced Simulation of Light Water Reactors (CASL) was established in July 2010 for the purpose of providing advanced modeling and simulation solutions for commercial nuclear reactors. The primary goal is to provide coupled, higher-fidelity, usable modeling and simulation capabilities than are currently available. These are needed to address light water reactor (LWR) operational and safety performance-defining phenomena that are not yet able to be fully modeled taking a first-principles approach. In order to pursue these goals, CASL has participation from laboratory, academic, and industry partners. These partners are pursuing the solution of ten major "Challenge Problems" in order to advance the state-of-the-art in reactor design and analysis to permit power uprates, higher burnup, life extension, and increased safety. At present, the problems being addressed by CASL are primarily reactor physics-oriented; however, this paper is intended to introduce CASL to the reactor dosimetry community because of the importance of reactor physics modelling and nuclear data to define the source term for that community and the applicability and extensibility of the transport methods being developed.

  5. Interfacing systems LOCA (loss-of-coolant accidents): Pressurized water reactors

    SciTech Connect

    Bozoki, G.; Kohut, P.; Fitzpatrick, R.

    1989-02-01

    This report summarizes a study performed by Brookhaven National Laboratory for the Office of Nuclear Regulatory Research, Reactor and Plant Safety Issues Branch, Division of Reactor and Plant Systems, US Nuclear Regulatory Commission. This study was requested by the NRC in order to provide a technical basis for the resolution of Generic Issue 105 ''Interfacing LOCA at LWRs.'' This report deals with pressurized water reactors (PWRs). A parallel report was also accomplished for boiling water reactors. This study focuses on three representative PWRs and extrapolates the plant-specific findings for their generic applicability. In addition, a generic analysis was performed to investigate the cost-benefit aspects of imposing a testing program that would require some minimum level of leak testing of the pressure isolation valves on plants that presently have no such requirements. 28 refs., 31 figs., 64 tabs.

  6. [Experimental research on combined water and air backwashing reactor technology for biological activated carbon].

    PubMed

    Xie, Zhi-Gang; Qiu, Xue-Min; Zhao, Yan-Ling

    2012-01-01

    To proper control the backwashing process of biological activated carbon (BAC) reactor and improve the overall operation performance, the evaluative indexes such as backwashing wastewater turbidity, organic pollutants removal rate of pre and post-backwashing, and the variation of biomass and biological activity in carbon column are used to compare and analyze the effect of three different combined water and air backwashing methods on the operation of BAC reactor. The result shows that intermittent combined water and air backwashing method is most suitable to BAC reactor. The biological activaty obviously increases by 62.5% after intermittent combined water and air backwashing process. While, the biological activaty using the backwashing method of air plus water and the backwashing method of water and air compounded plus water washing increases by 55.6%, 38.5%, respectively. After backwashing 308h, the reactor recovered to its normal function after intermittent combined water and air backwashing process with the removal rate of UV254 reaching to 60.0%. The fulvic-like fluorescence peak of backwashing water are very weak, and are characterized by low-excitation wavelength tryptophan like (peak S) and high excitation wavelength of tryptophan (peak T), which are caused by the microbial debris washed down. The three-dimensional fluorescence spectra also show that microbial fragments are easy to be washed clean with intermittent combined water and air backwashing.

  7. Kinetics of heavy metal inhibition of 1,2-dichloroethane biodegradation in co-contaminated water.

    PubMed

    Arjoon, Ashmita; Olaniran, Ademola Olufolahan; Pillay, Balakrishna

    2015-03-01

    Sites co-contaminated with heavy metals and 1,2-DCA may pose a greater challenge for bioremediation, as the heavy metals could inhibit the activities of microbes involved in biodegradation. Therefore, this study was undertaken to quantitatively assess the effects of heavy metals (arsenic, cadmium, mercury, and lead) on 1,2-DCA biodegradation in co-contaminated water. The minimum inhibitory concentrations (MICs) and concentrations of the heavy metals that caused half-life doubling (HLDs) of 1,2-DCA as well as the degradation rate coefficient (k(1)) and half-life (t(½)) of 1,2-DCA were measured and used to predict the toxicity of the heavy metals in the water microcosms. An increase in heavy metal concentration resulted in a progressive increase in the t(½) and relative t(½) and a decrease in k(1). The MICs and HLDs of the heavy metals were found to vary, depending on the heavy metals type. In addition, the presence of heavy metals was shown to inhibit 1,2-DCA biodegradation in a dose-dependent manner, with the following order of decreasing inhibitory effect: Hg(2+)  > As(3+)  > Cd(2+)  > Pb(2+). Findings from this study have significant implications for the development of bioremediation strategies for effective degradation of 1,2-DCA and other related compounds in wastewater co-contaminated with heavy metals.

  8. Physical characteristics of GE (General Electric) BWR (boiling-water reactor) fuel assemblies

    SciTech Connect

    Moore, R.S.; Notz, K.J.

    1989-06-01

    The physical characteristics of fuel assemblies manufactured by the General Electric Company for boiling-water reactors are classified and described. The classification into assembly types is based on the GE reactor product line, the Characteristics Data Base (CDB) assembly class, and the GE fuel design. Thirty production assembly types are identified. Detailed physical data are presented for each assembly type in an appendix. Descriptions of special (nonstandard) fuels are also reported. 52 refs., 1 fig., 6 tabs.

  9. Cask performance and interface specifications for shipment of US spent light water reactor fuel

    SciTech Connect

    Sanders, T.L.; Allen, G.C.; Wilmot, E.L.

    1986-01-01

    Preliminary cask physical performance specifications and interface guidelines have been defined to support the development of a family of casks and transporters for shipments in the United States (US) of spent light water reactor fuel. These shipments will be made from US commercial reactor facilities to high-level waste receiving facilities. The specified hardware consists of both truck and rail/barge casks along with their associated transporters.

  10. An effective means of biofiltration of heavy metal contaminated water bodies using aquatic weed Eichhornia crassipes.

    PubMed

    Tiwari, Suchi; Dixit, Savita; Verma, Neelam

    2007-06-01

    Various aquatic plant species are known to accumulate heavy metals through the process of bioaccumulation. World's most troublesome aquatic weed water hyacinth (Eichhornia crassipes) has been studied for its tendency to bio-accumulate and bio-magnify the heavy metal contaminants present in water bodies. The chemical investigation of plant parts has shown that it accumulates heavy metals like lead (Pb), chromium (Cr), zinc (Zn), manganese (Mn) and copper (Cu) to a large extent. Of all the heavy metals studied Pb, Zn and Mn tend to show greater affinity towards bioaccumulation. The higher concentration of metal in the aquatic weed signifies the biomagnification that lead to filtration of metallic ions from polluted water. The concept that E. crassipes can be used as a natural aquatic treatment system in the uptake of heavy metals is explored.

  11. IRIS Reactor a Suitable Option to Provide Energy and Water Desalination for the Mexican Northwest Region

    SciTech Connect

    Alonso, G.; Ramirez, R.; Gomez, C.; Viais, J.

    2004-10-03

    The Northwest region of Mexico has a deficit of potable water, along this necessity is the region growth, which requires of additional energy capacity. The IRIS reactor offers a very suitable source of energy given its modular size of 300 MWe and it can be coupled with a desalination plant to provide the potable water for human consumption, agriculture and industry. The present paper assess the water and energy requirements for the Northwest region of Mexico and how the deployment of the IRIS reactor can satisfy those necessities. The possible sites for deployment of Nuclear Reactors are considered given the seismic constraints and the closeness of the sea for external cooling. And in the other hand, the size of the desalination plant and the type of desalination process are assessed accordingly with the water deficit of the region.

  12. Directions for reactor target design based on the US heavy ion fusion systems assessment

    SciTech Connect

    Wilson, D.C.; Dudziak, D.; Magelssen, G.; Zuckerman, D.; Dreimeyer, D.

    1986-01-01

    We studied areas of major uncertainty in target design using the cost of electricity as our figure of merit. Net electric power from the plant was fixed at 1000 MW to eliminate large effects due to economies of scale. The system is relatively insensitive to target gain. Factors of three changes in gain cause only 8 to 12% changes in electricity cost. An increase in the peak power needed to drive targets poses only a small cost risk, but requires many more beamlets be transported to the target. A shortening of the required ion range causes both cost and beamlet difficulties. A factor of 4 decrease in the required range at a fixed driver energy increases electricity cost by 44% and raises the number of beamlets to 240. Finally, the heavy ion fusion system can accommodate large increases in target costs. To address the major uncertainties, target design should concentrate on the understanding requirements for ion range and peak driver power.

  13. Nuclear reactor with makeup water assist from residual heat removal system

    DOEpatents

    Corletti, Michael M.; Schulz, Terry L.

    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.

  14. 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.

  15. Reconstructing the direction of reactor antineutrinos via electron scattering in Gd-doped water Cherenkov detectors

    SciTech Connect

    Hellfeld, D.; Bernstein, A.; Dazeley, S.; Marianno, C.

    2017-01-01

    The potential of elastic antineutrino-electron scattering (ν¯e + e → ν¯e + e) in a Gd-doped water Cherenkov detector to determine the direction of a nuclear reactor antineutrino flux was investigated using the recently proposed WATCHMAN antineutrino experiment as a baseline model. The expected scattering rate was determined assuming a 13 km standoff from a 3.758 GWt light water nuclear reactor. Background was estimated via independent simulations and by appropriately scaling published measurements from similar detectors. Many potential backgrounds were considered, including solar neutrinos, misidentified reactor-based inverse beta decay interactions, cosmogenic radionuclide and water-borne radon decays, and gamma rays from the photomultiplier tubes, detector walls, and surrounding rock. The detector response was modeled using a GEANT4-based simulation package. The results indicate that with the use of low radioactivity PMTs and sufficient fiducialization, water-borne radon and cosmogenic radionuclides pose the largest threats to sensitivity. The directional sensitivity was then analyzed as a function of radon contamination, detector depth, and detector size. Lastly, the results provide a list of theoretical conditions that, if satisfied in practice, would enable nuclear reactor antineutrino directionality in a Gd-doped water Cherenkov detector approximately 10 km from a large power reactor.

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

    SciTech Connect

    Murphy, B.D.

    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 composition of the spent fuel is estimated at various times following discharge. Actinides and all significant fission products are considered. The activities, decay-heat values, and gamma-ray fluxes associated with the spent fuel are also discussed. It is clear from the analysis that following discharge the plutonium is no longer of weapons-grade composition. The characteristics of the mixed-oxide fuel at various times following discharge indicate its behavior under long-term storage. As a counterpoint to the mixed-oxide fuel case, the situation with a similar reactor fueled with uranium oxide alone is analyzed. The comparisons serve to emphasize the significance of the plutonium as part of the fuel. For the mixed-oxide case, the burnup was 42,200 MWd/MTHM; in the pure-uranium case, it was 47,800 MWd/MTHM.

  17. Experimental Study on Behavior of Bow-tie Tree Generation by Using Heavy Water

    NASA Astrophysics Data System (ADS)

    Kumazawa, Takao; Nakagawa, Wataru; Tsurumaru, Hidekazu

    Bow-tie tree (BTT) generated from contaminant, e.g., metal, carbon, amber(over cured resin) or void in insulator is a significant deterioration factor of XLPE power cable. However, essential role of water in generation and progress of BTT is not yet sufficiently cleared. In order to investigate the role of water we paid attention to difference in chemical properties of light water (H2O) and heavy water (D2O), moreover we evaluated influence of isotopic effect due to hydrogen and deuterium on behavior of BTT generation. In accelerated aging test the number of BTT in XLPE sample, in which copper powder of 500ppm was contaminated as BTT cores, dipped in heavy water (D2O:100wt%) decreased to one third compared with light water(D2O:0wt%). Furthermore, the maximum length of BTT decreased with increase in concentration of heavy water. The experimental results show that heavy water exerted a depression effect on generation and progress of BTT. We considered that the depression effect due to hydrogen isotope appeared by inhibiting ionization and elution of BTT cores, because salt-solubility and ionic mobility of heavy water are about 15 to 20% smaller than those of light water. Therefore, the essential role of water seemed to be production and transport of ions in XLPE.

  18. 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

    ... COMMISSION Cost-Benefit Analysis for Radwaste Systems for Light-Water-Cooled Nuclear Power Reactors AGENCY... Systems for Light-Water-Cooled Nuclear Power Reactors,'' in which the NRC made editorial corrections and... analysis for liquid and gaseous radwaste system components for light water nuclear power...

  19. Seventh international symposium on environmental degradation of materials in nuclear power systems -- Water reactors: Proceedings and symposium discussions. Volume 2

    SciTech Connect

    Airey, G.; Andresen, P.; Brown, J.

    1995-12-31

    The papers in this volume are divided into the following sections: boiling water reactors and pressurized water reactors -- water chemistry; radiation effects; modelling and life prediction; irradiation assisted stress corrosion cracking; pressure vessel and low alloy steels; and fuel cladding/wear. Separate abstracts were prepared for most papers in this volume.

  20. Heavy metal contamination of soil and water in the vicinity of an abandoned e-waste recycling site: implications for dissemination of heavy metals.

    PubMed

    Wu, Qihang; Leung, Jonathan Y S; Geng, Xinhua; Chen, Shejun; Huang, Xuexia; Li, Haiyan; Huang, Zhuying; Zhu, Libin; Chen, Jiahao; Lu, Yayin

    2015-02-15

    Illegal e-waste recycling activity has caused heavy metal pollution in many developing countries, including China. In recent years, the Chinese government has strengthened enforcement to impede such activity; however, the heavy metals remaining in the abandoned e-waste recycling site can still pose ecological risk. The present study aimed to investigate the concentrations of heavy metals in soil and water in the vicinity of an abandoned e-waste recycling site in Longtang, South China. Results showed that the surface soil of the former burning and acid-leaching sites was still heavily contaminated with Cd (>0.39 mg kg(-1)) and Cu (>1981 mg kg(-1)), which exceeded their respective guideline levels. The concentration of heavy metals generally decreased with depth in both burning site and paddy field, which is related to the elevated pH and reduced TOM along the depth gradient. The pond water was seriously acidified and contaminated with heavy metals, while the well water was slightly contaminated since heavy metals were mostly retained in the surface soil. The use of pond water for irrigation resulted in considerable heavy metal contamination in the paddy soil. Compared with previous studies, the reduced heavy metal concentrations in the surface soil imply that heavy metals were transported to the other areas, such as pond. Therefore, immediate remediation of the contaminated soil and water is necessary to prevent dissemination of heavy metals and potential ecological disaster. Copyright © 2014 Elsevier B.V. All rights reserved.

  1. Evolutionary/advanced light water reactor data report

    SciTech Connect

    1996-02-09

    The US DOE Office of Fissile Material Disposition is examining options for placing fissile materials that were produced for fabrication of weapons, and now are deemed to be surplus, into a condition that is substantially irreversible and makes its use in weapons inherently more difficult. The principal fissile materials subject to this disposition activity are plutonium and uranium containing substantial fractions of plutonium-239 uranium-235. The data in this report, prepared as technical input to the fissile material disposition Programmatic Environmental Impact Statement (PEIS) deal only with the disposition of plutonium that contains well over 80% plutonium-239. In fact, the data were developed on the basis of weapon-grade plutonium which contains, typically, 93.6% plutonium-239 and 5.9% plutonium-240 as the principal isotopes. One of the options for disposition of weapon-grade plutonium being considered is the power reactor alternative. Plutonium would be fabricated into mixed oxide (MOX) fuel and fissioned (``burned``) in a reactor to produce electric power. The MOX fuel will contain dioxides of uranium and plutonium with less than 7% weapon-grade plutonium and uranium that has about 0.2% uranium-235. The disposition mission could, for example, be carried out in existing power reactors, of which there are over 100 in the United States. Alternatively, new LWRs could be constructed especially for disposition of plutonium. These would be of the latest US design(s) incorporating numerous design simplifications and safety enhancements. These ``evolutionary`` or ``advanced`` designs would offer not only technological advances, but also flexibility in siting and the option of either government or private (e.g., utility) ownership. The new reactor designs can accommodate somewhat higher plutonium throughputs. This data report deals solely with the ``evolutionary`` LWR alternative.

  2. Guidance for Developing Principal Design Criteria for Advanced (Non-Light Water) Reactors

    SciTech Connect

    Holbrook, Mark; Kinsey, Jim

    2015-03-01

    In July 2013, the US Department of Energy (DOE) and US Nuclear Regulatory Commission (NRC) established a joint initiative to address a key portion of the licensing framework essential to advanced (non-light water) reactor technologies. The initiative addressed the “General Design Criteria for Nuclear Power Plants,” Appendix A to10 Code of Federal Regulations (CFR) 50, which were developed primarily for light water reactors (LWRs), specific to the needs of advanced reactor design and licensing. The need for General Design Criteria (GDC) clarifications in non-LWR applications has been consistently identified as a concern by the industry and varied stakeholders and was acknowledged by the NRC staff in their 2012 Report to Congress1 as an area for enhancement. The initiative to adapt GDC requirements for non-light water advanced reactor applications is being accomplished in two phases. Phase 1, managed by DOE, consisted of reviews, analyses and evaluations resulting in recommendations and deliverables to NRC as input for NRC staff development of regulatory guidance. Idaho National Laboratory (INL) developed this technical report using technical and reactor technology stakeholder inputs coupled with analysis and evaluations provided by a team of knowledgeable DOE national laboratory personnel with input from individual industry licensing consultants. The DOE national laboratory team reviewed six different classes of emerging commercial reactor technologies against 10 CFR 50 Appendix A GDC requirements and proposed guidance for their adapted use in non-LWR applications. The results of the Phase 1 analysis are contained in this report. A set of draft Advanced Reactor Design Criteria (ARDC) has been proposed for consideration by the NRC in the establishment of guidance for use by non-LWR designers and NRC staff. The proposed criteria were developed to preserve the underlying safety bases expressed by the original GDC, and recognizing that advanced reactors may take

  3. Microstructure evolution and degradation mechanisms of reactor internal steel irradiated with heavy ions

    NASA Astrophysics Data System (ADS)

    Borodin, O. V.; Bryk, V. V.; Kalchenko, A. S.; Parkhomenko, A. A.; Shilyaev, B. A.; Tolstolutskaya, G. D.; Voyevodin, V. N.

    2009-03-01

    Structure evolution and degradation mechanisms during irradiation of 18Cr-10Ni-Ti steel (material of VVER-1000 reactor internals are investigated). Using accelerator irradiations with Cr3+ and Ar+ ions allowed studying effects of dose rate, different initial structure state and implanted ions on features of structure evolution and main mechanisms of degradation including low temperature swelling and embrittlement of the 18Cr-10Ni-Ti steel. It is shown that differences in dose rate at most irradiation temperatures mainly exert their influence on the duration of the swelling transient regime. Calculations of possible transmutation products during irradiation of this steel in a VVER-1000 spectrum were performed. It is shown that gaseous atoms (He and H), which are generated simultaneously with radiation defects, stabilize the elements of radiation microstructure and influence the swelling. The nature of deformation under different temperatures of irradiation and of mechanical testing is investigated. It is shown that the temperature sensitivity of swelling behaviour in the investigated steel, with different initial structures can be connected with the dynamic behaviour of point defect sinks.

  4. Lactose Variability of Escherichia coli in Thermally Stressed Reactor Effluent Waters

    PubMed Central

    Kasweck, K. L.; Fliermans, C. B.

    1978-01-01

    Lactose-utilizing and nalidixic acid-resistant populations of Escherichia coli, having an optimum growth temperature of 37°C, were placed in modified diffusion chambers. The chambers were submerged in the epilimnion and hypolimnion of a 1,100-hectare lake (Par Pond) which receives cooling water from a nuclear production reactor. Control chambers were placed in a deep-water reservoir and a Flowing-Streams Laboratory, both of which had comparable temperatures to Par Pond. The populations of E. coli were sampled regularly for up to 3 weeks. Viability of the bacteria was determined by dilution plating to nutrient agar followed by replicate plating onto selective medium to determine lactose utilization and nalidixic acid sensitivity. Initial populations of E. coli were lactose positive but changed to lactose negative in Par Pond when the reactor was operating (i.e., cooling water from the heat exchangers was being discharged to the lake). This alteration occurred most rapidly in the chambers closest to the cooling-water discharge point. Such changes did not occur in a deep-water reservoir, in Par Pond when the reactor was not operating, or in the Flowing-Streams Laboratory. The nalidixic acid-resistant characteristic remained stable regardless of the chambers' placement or reactor operations. Although the reasons for such alterations are unclear, it appears that lactose-negative populations of E. coli are selected for in these reactor effluent waters. The loss of the lactose characteristic prevents the recognition and identification of E. coli in this cooling lake (when the reactor is operating) and may prevent the assessment of water quality based on coliform recognition. PMID:365111

  5. The ISS Water Processor Catalytic Reactor as a Post Processor for Advanced Water Reclamation Systems

    NASA Technical Reports Server (NTRS)

    Nalette, Tim; Snowdon, Doug; Pickering, Karen D.; Callahan, Michael

    2007-01-01

    Advanced water processors being developed for NASA s Exploration Initiative rely on phase change technologies and/or biological processes as the primary means of water reclamation. As a result of the phase change, volatile compounds will also be transported into the distillate product stream. The catalytic reactor assembly used in the International Space Station (ISS) water processor assembly, referred to as Volatile Removal Assembly (VRA), has demonstrated high efficiency oxidation of many of these volatile contaminants, such as low molecular weight alcohols and acetic acid, and is considered a viable post treatment system for all advanced water processors. To support this investigation, two ersatz solutions were defined to be used for further evaluation of the VRA. The first solution was developed as part of an internal research and development project at Hamilton Sundstrand (HS) and is based primarily on ISS experience related to the development of the VRA. The second ersatz solution was defined by NASA in support of a study contract to Hamilton Sundstrand to evaluate the VRA as a potential post processor for the Cascade Distillation system being developed by Honeywell. This second ersatz solution contains several low molecular weight alcohols, organic acids, and several inorganic species. A range of residence times, oxygen concentrations and operating temperatures have been studied with both ersatz solutions to provide addition performance capability of the VRA catalyst.

  6. Nonisothermal reactors for the production of pure water from peritoneal dialysis waste waters.

    PubMed

    Diano, N; Ettari, G; Grano, V; Gaeta, F S; Rossi, S; Bencivenga, U; D'Alterio, C; Ruocco, G; Mita, L; De Santo, N G; Canciglia, P; Mita, D G

    2007-01-01

    The diffusion of peritoneal dialysis (PD) at home is somewhat restricted by the difficulty of transport and storage of a large amount of dialytic solutions. This problem is exacerbated in the case of hemodialysis. With the aim of producing pure water to be used in preparing the solution for peritoneal dialysis, or for hemodialysis in general, as one example, we purified the spent dialysate solution from PD. Experiments were carried out with 24 dialysate solutions taken from 8 patients. Pure water was obtained by means of a thermodialysis process in a hollow fiber reactor operating under nonisothermal conditions. Results show that the yield of the nonisothermal process is dependent on the temperature difference applied across the hydrophobic membranes. The production of pure water per square meter of membrane and per hour was equal to 0.55 or 1.2 or 2.0 liters, with a temperature difference of 11 degrees C or 21 degrees C or 28 degrees C, respectively. These results encourage the use of the thermodialysis process in the production of pure water for clinical uses.

  7. Assessment of Possible Cycle Lengths for Fully-Ceramic Micro-Encapsulated Fuel-Based Light Water Reactor Concepts

    SciTech Connect

    R. Sonat Sen; Michael A. Pope; Abderrafi M. Ougouag; Kemal O. Pasamehmetoglu

    2012-04-01

    The tri-isotropic (TRISO) fuel developed for High Temperature reactors is known for its extraordinary fission product retention capabilities [1]. Recently, the possibility of extending the use of TRISO particle fuel to Light Water Reactor (LWR) technology, and perhaps other reactor concepts, has received significant attention [2]. The Deep Burn project [3] currently focuses on once-through burning of transuranic fissile and fissionable isotopes (TRU) in LWRs. The fuel form for this purpose is called Fully-Ceramic Micro-encapsulated (FCM) fuel, a concept that borrows the TRISO fuel particle design from high temperature reactor technology, but uses SiC as a matrix material rather than graphite. In addition, FCM fuel may also use a cladding made of a variety of possible material, again including SiC as an admissible choice. The FCM fuel used in the Deep Burn (DB) project showed promising results in terms of fission product retention at high burnup values and during high-temperature transients. In the case of DB applications, the fuel loading within a TRISO particle is constituted entirely of fissile or fissionable isotopes. Consequently, the fuel was shown to be capable of achieving reasonable burnup levels and cycle lengths, especially in the case of mixed cores (with coexisting DB and regular LWR UO2 fuels). In contrast, as shown below, the use of UO2-only FCM fuel in a LWR results in considerably shorter cycle length when compared to current-generation ordinary LWR designs. Indeed, the constraint of limited space availability for heavy metal loading within the TRISO particles of FCM fuel and the constraint of low (i.e., below 20 w/0) 235U enrichment combine to result in shorter cycle lengths compared to ordinary LWRs if typical LWR power densities are also assumed and if typical TRISO particle dimensions and UO2 kernels are specified. The primary focus of this summary is on using TRISO particles with up to 20 w/0 enriched uranium kernels loaded in Pressurized Water

  8. Thermalhydraulic calculation for boiling water reactor and its natural circulation component

    SciTech Connect

    Trianti, Nuri Nurjanah,; Su’ud, Zaki; Arif, Idam; Permana, Sidik

    2015-09-30

    Thermalhydraulic of reactor core is the thermal study on fluids within the core reactor, i.e. analysis of the thermal energy transfer process produced by fission reaction from fuel to the reactor coolant. This study include of coolant temperature and reactor power density distribution. The purposes of this analysis in the design of nuclear power plant are to calculate the coolant temperature distribution and the chimney height so natural circulation could be occurred. This study was used boiling water reactor (BWR) with cylinder type reactor core. Several reactor core properties such as linear power density, mass flow rate, coolant density and inlet temperature has been took into account to obtain distribution of coolant density, flow rate and pressure drop. The results of calculation are as follows. Thermal hydraulic calculations provide the uniform pressure drop of 1.1 bar for each channels. The optimum mass flow rate to obtain the uniform pressure drop is 217g/s. Furthermore, from the calculation it could be known that outlet temperature is 288°C which is the saturated fluid’s temperature within the system. The optimum chimney height for natural circulation within the system is 14.88 m.

  9. "Periodic-table-style" paper device for monitoring heavy metals in water.

    PubMed

    Li, Miaosi; Cao, Rong; Nilghaz, Azadeh; Guan, Liyun; Zhang, Xiwang; Shen, Wei

    2015-03-03

    If a paper-based analytical device (μ-PAD) could be made by printing indicators for detection of heavy metals in chemical symbols of the metals in a style of the periodic table of elements, it could be possible for such μ-PAD to report the presence and the safety level of heavy metal ions in water simultaneously and by text message. This device would be able to provide easy solutions to field-based monitoring of heavy metals in industrial wastewater discharges and in irrigating and drinking water. Text-reporting could promptly inform even nonprofessional users of the water quality. This work presents a proof of concept study of this idea. Cu(II), Ni(II), and Cr(VI) were chosen to demonstrate the feasibility, specificity, and reliability of paper-based text-reporting devices for monitoring heavy metals in water.

  10. Neutron collar calibration for assay of LWR (light-water reactor) fuel assemblies

    SciTech Connect

    Menlove, H.O.; Pieper, J.E.

    1987-03-01

    The neutron-coincidence collar is used for the verification of the uranium content in light-water reactor fuel assemblies. An AmLi neutron source is used to give an active interrogation of the fuel assembly to measure the /sup 235/U content, and the /sup 238/U content is verified from a passive neutron-coincidence measurement. This report gives the collar calibration data of pressurized-water reactor and boiling-water reactor fuel assemblies. Calibration curves and correction factors are presented for neutron absorbers (burnable poisons) and different fuel assembly sizes. The data were collected at Exxon Nuclear, Franco-Belge de Fabrication de Combustibles, ASEA-Atom, and other nuclear fuel fabrication facilities.

  11. Heavy metal bioaccumulation and effects on water hyacinth weevils, Neochetina eichhorniae, feeding on water hyacinth, Eichhornia crassipes

    SciTech Connect

    Kay, S.H.; Haller, W.T.

    1986-08-01

    Both aquatic and terrestrial habitats frequently are subject to contamination by toxic heavy metals, yet very little is known about the influence of heavy metals absorbed by plant tissues upon the phytophagous insect fauna feeding upon these plants. The objectives of this study were to determine the influence of plant-absorbed metals upon the feeding, mortality, and body burdens of lead, cadmium, and copper in the water hyacinth weevil, Neochetina eichhorniae, imported for the biological control of water hyacinths (Eichhornia crassipes).

  12. Zirconium carbide coating for corium experiments related to water-cooled and sodium-cooled reactors

    NASA Astrophysics Data System (ADS)

    Plevacova, K.; Journeau, C.; Piluso, P.; Zhdanov, V.; Baklanov, V.; Poirier, J.

    2011-07-01

    Since the TMI and Chernobyl accidents the risk of nuclear severe accident is intensively studied for existing and future reactors. In case of a core melt-down accident in a nuclear reactor, a complex melt, called corium, forms. To be able to perform experiments with prototypic corium materials at high temperature, a coating which resists to different corium melts related to Generation I and II Water Reactors and Generation IV sodium fast reactor was researched in our experimental platforms both in IAE NNC in Kazakhstan and in CEA in France. Zirconium carbide was selected as protective coating for graphite crucibles used in our induction furnaces: VCG-135 and VITI. The method of coating application, called reactive wetting, was developed. Zirconium carbide revealed to resist well to the (U x, Zr y)O 2-z water reactor corium. It has also the advantage not to bring new elements to this chemical system. The coating was then tested with sodium fast reactor corium melts containing steel or absorbers. Undesirable interactions were observed between the coating and these materials, leading to the carburization of the corium ingots. Concerning the resistance of the coating to oxide melts without ZrO 2, the zirconium carbide coating keeps its role of protective barrier with UO 2-Al 2O 3 below 2000 °C but does not resist to a UO 2-Eu 2O 3 mixture.

  13. Bioremediation of chlorobenzene-contaminated ground water in an in situ reactor mediated by hydrogen peroxide.

    PubMed

    Vogt, Carsten; Alfreider, Albin; Lorbeer, Helmut; Hoffmann, Doreen; Wuensche, Lothar; Babel, Wolfgang

    2004-01-01

    New in situ reactive barrier technologies were tested nearby a local aquifer in Bitterfeld, Saxonia-Anhalt, Germany, which is polluted mainly by chlorobenzene (CB), in concentrations up to 450 microM. A reactor filled with original aquifer sediment was designed for the microbiological remediation of the ground water by indigenous bacterial communities. Two remediation variants were examined: (a) the degradation of CB under anoxic conditions in the presence of nitrate; (b) the degradation of CB under mixed electron acceptor conditions (oxygen+nitrate) using hydrogen peroxide as the oxygen-releasing compound. Under anoxic conditions, no definite degradation of CB was observed. Adding hydrogen peroxide (2.94 mM) and nitrate (2 mM) led to the disappearance of CB (ca. 150 microM) in the lower part of the reactor, accompanied by a strong increase of the number of cultivable aerobic CB degrading bacteria in reactor water and sediment samples, indicating that CB was degraded mainly by productive bacterial metabolism. Several aerobic CB degrading bacteria, mostly belonging to the genera Pseudomonas and Rhodococcus, were isolated from reactor water and sediments. In laboratory experiments with reactor water, oxygen was rapidly released by hydrogen peroxide, whereas biotic-induced decomposition reactions of hydrogen peroxide were almost four times faster than abiotic-induced decomposition reactions. A clear chemical degradation of CB mediated by hydrogen peroxide was not observed. CB was also completely degraded in the reactor after reducing the hydrogen peroxide concentration to 880 microM. The CB degradation completely collapsed after reducing the hydrogen peroxide concentration to 440 microM. In the following, the hydrogen peroxide concentrations were increased again (to 880 microM, 2.94 mM, and 880 microM, respectively), but the oxygen demand for CB degradation was higher than observed before, indicating a shift in the bacterial population. During the whole experiment

  14. Accumulation of heavy metals in water, sediments and wetland plants of kizilirmak delta (samsun, Turkey).

    PubMed

    Engin, M S; Uyanik, A; Kutbay, H G

    2015-01-01

    In this study, concentrations of heavy metals (Fe, Mn, Ni, Co, Zn, Cu, and Pb) were measured in water bodies including streams, bottom sediments and various wetland plants of Kızılırmak Delta. Kızılırmak Delta is one of the largest and the most important natural wetlands in Turkey and has been protected by Ramsar convention since 1993. The heavy metal concentrations in water were found lower than that of national standards for protected lakes and reserves. In bottom sediments and wetland plants, however, the accumulated amounts of different heavy metals varied in the following order: Fe>Mn>Zn>Ni>Co>Cu>Pb, and Fe>Mn>Zn>Ni>Co respectively. Heavy metal uptake of Hydrocharis morsus-ranae and Myriophyllum verticillatum plants among others were found far above the toxic levels and they might be used as bio-indicators and heavy metal accumulators in polluted natural areas.

  15. [Heavy metals distribution characteristics and risk assessment of water below an electroplating factory].

    PubMed

    Hang, Xiao-Shuai; Wang, Huo-Yan; Zhou, Jian-Min

    2008-10-01

    Surface water and shallow groundwater within the flow of an electroplating factory was analyzed in order to study the resulting impact. The analysis method of ICP-AES was used to analyze content of zinc, manganese, chromium, copper and nickel in surface water and groundwater samples. The results indicate acidic pollutants of zinc, manganese, chromium, copper and nickel were discharged from the factory with concentrations of 1.34, 3.77, 28.1, 6.40 and 9.37 mg x L(-1), respectively; and pH was 2.32. They all exceeded permissible levels according to Integrated Wastewater Discharge Standard except zinc. Factory discharge is responsible for the longitudinal distribution characteristics of heavy metals in the stream water downstream from the factory. Heavy metals variations in the well water do not suggest they were affected by heavy metals in the stream, indicating that the migration rates of heavy metals in soils were relatively low. Risk assessment shows surface water quality significantly deteriorated. Nickel and manganese in the stream water exceeded the standard levels seriously, and chromium and copper in some samples were also above Grade III standard levels according to Environmental Quality Standard for Surface Water. Moreover, all studied heavy metals in 14 groundwater samples measured within drinking water standard, except manganese in 4 groundwater samples, which were Grade IV according to Quality Standard for Ground water.

  16. Analysis of Pressurized Water Reactor Primary Coolant Leak Events Caused by Thermal Fatigue

    SciTech Connect

    C. L. Atwood; V. N. Shah; W. J. Galyean

    1999-09-01

    We present statistical analyses of pressurized water reactor (PWR) primary coolant leak events caused by thermal fatigue, and discuss their safety significance. Our worldwide data contain 13 leak events (through-wall cracking) in 3509 reactor-years, all in stainless steel piping with diameter less than 25 cm. Several types of data analysis show that the frequency of leak events (events per reactor-year) is increasing with plant age, and the increase is statistically significant. When an exponential trend model is assumed, the leak frequency is estimated to double every 8 years of reactor age, although this result should not be extrapolated to plants much older than 25 years. Difficulties in arresting this increase include lack of quantitative understanding of the phenomena causing thermal fatigue, lack of understanding of crack growth, and difficulty in detecting existing cracks.

  17. Tritium recovery from tritiated water with a two-stage palladium membrane reactor

    SciTech Connect

    Birdsell, S.A.; Willms, R.S.

    1997-04-01

    A process to recover tritium from tritiated water has been successfully demonstrated at TSTA. The 2-stage palladium membrane reactor (PMR) is capable of recovering tritium from water without generating additional waste. This device can be used to recover tritium from the substantial amount of tritiated water that is expected to be generated in the International Thermonuclear Experimental Reactor both from torus exhaust and auxiliary operations. A large quantity of tritiated waste water exists world wide because the predominant method of cleaning up tritiated streams is to oxidize tritium to tritiated water. The latter can be collected with high efficiency for subsequent disposal. The PMR is a combined catalytic reactor/permeator. Cold (non-tritium) water processing experiments were run in preparation for the tritiated water processing tests. Tritium was recovered from a container of molecular sieve loaded with 2,050 g (2,550 std. L) of water and 4.5 g of tritium. During this experiment, 27% (694 std. L) of the water was processed resulting in recovery of 1.2 g of tritium. The maximum water processing rate for the PMR system used was determined to be 0.5 slpm. This correlates well with the maximum processing rate determined from the smaller PMR system on the cold test bench and has resulted in valuable scale-up and design information.

  18. Membrane chemical reactor (MCR) combining photocatalysis and microfiltration for grey water treatment.

    PubMed

    Rivero, M J; Parsons, S A; Jeffrey, P; Pidou, M; Jefferson, B

    2006-01-01

    Urban water recycling is now becoming an important issue where water resources are becoming scarce. This paper looks at reusing grey water; the preference is treatment processes based on biological systems to remove the dissolved organic content. Here, an alternative process, photocatalysis is discussed as it is an attractive technology that could be well-suited for treating the recalcitrant organic compounds found in grey water. The photocatalytic process oxidises organic reactants at a catalyst surface in the presence of ultraviolet light. Given enough exposure time, organic compounds will be oxidized into CO2 and water. The best contact is achieved in a slurry reactor but a second step to separate and recover the catalyst is need. This paper discusses a new membrane chemical reactor (MCR) combining photocatalysis and microfiltration for grey water treatment.

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

    SciTech Connect

    Forseberg, C.W.

    1990-01-01

    This document describes passive decay-heat removal system for a water-cooled nuclear reactor which employs a closed heat transfer loop having heat-exchanging coils inside an open-topped, insulated evaporator 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.

  20. Black water sludge reuse in agriculture: are heavy metals a problem?

    PubMed

    Tervahauta, Taina; Rani, Sonia; Hernández Leal, Lucía; Buisman, Cees J N; Zeeman, Grietje

    2014-06-15

    Heavy metal content of sewage sludge is currently the most significant factor limiting its reuse in agriculture within the European Union. In the Netherlands most of the produced sewage sludge is incinerated, mineralizing the organic carbon into the atmosphere rather than returning it back to the soil. Source-separation of black water (toilet water) excludes external heavy metal inputs, such as industrial effluents and surface run-offs, producing sludge with reduced heavy metal content that is a more favorable source for resource recovery. The results presented in this paper show that feces is the main contributor to the heavy metal loading of vacuum collected black water (52-84%), while in sewage the contribution of feces is less than 10%. To distinguish black water from sewage in the sludge reuse regulation, a control parameter should be implemented, such as the Hg and Pb content that is significantly higher in sewage sludge compared to black water sludge (from 50- to 200-fold). The heavy metals in feces and urine are primarily from dietary sources, and promotion of the soil application of black water sludge over livestock manure and artificial fertilizers could further reduce the heavy metal content in the soil/food cycle.

  1. Design and Testing of Vacuum Breaker Check Valve for Simplified Boiling Water Reactor

    SciTech Connect

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

    2002-07-01

    A new design of the vacuum breaker check valve was developed to replace the mechanical valve in a simplified boiling water reactor. Scaling and design calculations were performed to obtain the geometry of new passive hydraulic vacuum breaker check valve. In order to check the valve performance, a RELAP5 model of the simplified boiling water reactor system with the new valve was developed. The valve was implemented in an integral facility, PUMA and was tested for large break loss of coolant accident. (authors)

  2. Flow-induced vibration for light-water reactors. Progress report, April 1978-December 1979

    SciTech Connect

    Schardt, J. F.

    1980-03-01

    Flow-Induced vibration for Light Water Reactors (FIV for LWRs) is a four-year program designed to improve the FIV performance of light water reactors through the development of design criteria, analytical models for predicting behavior of components, general scaling laws to improve the accuracy of reduced-scale tests, and the identification of high FIV risk areas. The program commenced December 1, 1976, but was suspended on September 30, 1978, due to a shift in Department of Energy (DOE) priorities away from LWR productivity/availability. It was reinitiated as of August 1, 1979. This progress report summarizes the accomplishments achieved during the period from April 1978 to December 1979.

  3. Nuclear Thermal Propulsion engine based on Particle Bed Reactor using light water steam as a propellant

    SciTech Connect

    Powell, J.R.; Ludewig, H.; Maise, G.

    1993-06-01

    In this paper the possibility of configuring a water cooled Nuclear Thermal Propulsion (NTP) rocket, based on a Particle Bed Reactor (PBR) is investigated. This rocket will be used to operate on water obtained from near earth objects. The conclusions reached in this paper indicate that it is possible to configure a PBR based NTP rocket to operate on water and meet the mission requirements envisioned for it. No insurmountable technology issues have been identified.

  4. Light water reactor mixed-oxide fuel irradiation experiment

    SciTech Connect

    Hodge, S.A.; Cowell, B.S.; Chang, G.S.; Ryskamp, J.M.

    1998-06-01

    The United States Department of Energy Office of Fissile Materials Disposition is sponsoring and Oak Ridge National Laboratory (ORNL) is leading an irradiation experiment to test mixed uranium-plutonium oxide (MOX) fuel made from weapons-grade (WG) plutonium. In this multiyear program, sealed capsules containing MOX fuel pellets fabricated at Los Alamos National Laboratory (LANL) are being irradiated in the Advanced Test Reactor (ATR) at the Idaho National Engineering and Environmental Laboratory (INEEL). The planned experiments will investigate the utilization of dry-processed plutonium, the effects of WG plutonium isotopics on MOX performance, and any material interactions of gallium with Zircaloy cladding.

  5. Comparative study of electrical breakdown properties of deionized water and heavy water under pulsed power conditions

    SciTech Connect

    Veda Prakash, G.; Kumar, R.; Saurabh, K.; Nasir,; Anitha, V. P.; Chowdhuri, M. B.; Shyam, A.

    2016-01-15

    A comparative study of electrical breakdown properties of deionized water (H{sub 2}O) and heavy water (D{sub 2}O) is presented with two different electrode materials (stainless steel (SS) and brass) and polarity (positive and negative) combinations. The pulsed (∼a few tens of nanoseconds) discharges are conducted by applying high voltage (∼a few hundred kV) pulse between two hemisphere electrodes of the same material, spaced 3 mm apart, at room temperature (∼26-28 °C) with the help of Tesla based pulse generator. It is observed that breakdown occurred in heavy water at lesser voltage and in short duration compared to deionized water irrespective of the electrode material and applied voltage polarity chosen. SS electrodes are seen to perform better in terms of the voltage withstanding capacity of the liquid dielectric as compared to brass electrodes. Further, discharges with negative polarity are found to give slightly enhanced discharge breakdown voltage when compared with those with positive polarity. The observations corroborate well with conductivity measurements carried out on original and post-treated liquid samples. An interpretation of the observations is attempted using Fourier transform infrared measurements on original and post-treated liquids as well as in situ emission spectra studies. A yet another important observation from the emission spectra has been that even short (nanosecond) duration discharges result in the formation of a considerable amount of ions injected into the liquid from the electrodes in a similar manner as reported for long (microseconds) discharges. The experimental observations show that deionised water is better suited for high voltage applications and also offer a comparison of the discharge behaviour with different electrodes and polarities.

  6. Comparative study of electrical breakdown properties of deionized water and heavy water under pulsed power conditions.

    PubMed

    Veda Prakash, G; Kumar, R; Saurabh, K; Nasir; Anitha, V P; Chowdhuri, M B; Shyam, A

    2016-01-01

    A comparative study of electrical breakdown properties of deionized water (H2O) and heavy water (D2O) is presented with two different electrode materials (stainless steel (SS) and brass) and polarity (positive and negative) combinations. The pulsed (∼a few tens of nanoseconds) discharges are conducted by applying high voltage (∼a few hundred kV) pulse between two hemisphere electrodes of the same material, spaced 3 mm apart, at room temperature (∼26-28 °C) with the help of Tesla based pulse generator. It is observed that breakdown occurred in heavy water at lesser voltage and in short duration compared to deionized water irrespective of the electrode material and applied voltage polarity chosen. SS electrodes are seen to perform better in terms of the voltage withstanding capacity of the liquid dielectric as compared to brass electrodes. Further, discharges with negative polarity are found to give slightly enhanced discharge breakdown voltage when compared with those with positive polarity. The observations corroborate well with conductivity measurements carried out on original and post-treated liquid samples. An interpretation of the observations is attempted using Fourier transform infrared measurements on original and post-treated liquids as well as in situ emission spectra studies. A yet another important observation from the emission spectra has been that even short (nanosecond) duration discharges result in the formation of a considerable amount of ions injected into the liquid from the electrodes in a similar manner as reported for long (microseconds) discharges. The experimental observations show that deionised water is better suited for high voltage applications and also offer a comparison of the discharge behaviour with different electrodes and polarities.

  7. Bacterial Colonization of Pellet Softening Reactors Used during Drinking Water Treatment▿

    PubMed Central

    Hammes, Frederik; Boon, Nico; Vital, Marius; Ross, Petra; Magic-Knezev, Aleksandra; Dignum, Marco

    2011-01-01

    Pellet softening reactors are used in centralized and decentralized drinking water treatment plants for the removal of calcium (hardness) through chemically induced precipitation of calcite. This is accomplished in fluidized pellet reactors, where a strong base is added to the influent to increase the pH and facilitate the process of precipitation on an added seeding material. Here we describe for the first time the opportunistic bacterial colonization of the calcite pellets in a full-scale pellet softening reactor and the functional contribution of these colonizing bacteria to the overall drinking water treatment process. ATP analysis, advanced microscopy, and community fingerprinting with denaturing gradient gel electrophoretic (DGGE) analysis were used to characterize the biomass on the pellets, while assimilable organic carbon (AOC), dissolved organic carbon, and flow cytometric analysis were used to characterize the impact of the biological processes on drinking water quality. The data revealed pellet colonization at concentrations in excess of 500 ng of ATP/g of pellet and reactor biomass concentrations as high as 220 mg of ATP/m3 of reactor, comprising a wide variety of different microorganisms. These organisms removed as much as 60% of AOC from the water during treatment, thus contributing toward the biological stabilization of the drinking water. Notably, only a small fraction (about 60,000 cells/ml) of the bacteria in the reactors was released into the effluent under normal conditions, while the majority of the bacteria colonizing the pellets were captured in the calcite structures of the pellets and were removed as a reusable product. PMID:21148700

  8. Bacterial colonization of pellet softening reactors used during drinking water treatment.

    PubMed

    Hammes, Frederik; Boon, Nico; Vital, Marius; Ross, Petra; Magic-Knezev, Aleksandra; Dignum, Marco

    2011-02-01

    Pellet softening reactors are used in centralized and decentralized drinking water treatment plants for the removal of calcium (hardness) through chemically induced precipitation of calcite. This is accomplished in fluidized pellet reactors, where a strong base is added to the influent to increase the pH and facilitate the process of precipitation on an added seeding material. Here we describe for the first time the opportunistic bacterial colonization of the calcite pellets in a full-scale pellet softening reactor and the functional contribution of these colonizing bacteria to the overall drinking water treatment process. ATP analysis, advanced microscopy, and community fingerprinting with denaturing gradient gel electrophoretic (DGGE) analysis were used to characterize the biomass on the pellets, while assimilable organic carbon (AOC), dissolved organic carbon, and flow cytometric analysis were used to characterize the impact of the biological processes on drinking water quality. The data revealed pellet colonization at concentrations in excess of 500 ng of ATP/g of pellet and reactor biomass concentrations as high as 220 mg of ATP/m(3) of reactor, comprising a wide variety of different microorganisms. These organisms removed as much as 60% of AOC from the water during treatment, thus contributing toward the biological stabilization of the drinking water. Notably, only a small fraction (about 60,000 cells/ml) of the bacteria in the reactors was released into the effluent under normal conditions, while the majority of the bacteria colonizing the pellets were captured in the calcite structures of the pellets and were removed as a reusable product.

  9. Heavy metals in natural waters: applied monitoring and impact assessment

    SciTech Connect

    Moore, J.W.; Ramamoorthy, S.

    1984-01-01

    An attempt has been made to review a large amount of environmental data on eight common heavy metals. The chemistry, uptake and toxicity of arsenic, cadmium, chromium, copper, lead, mercury, nickel, and zinc are presented. With this information, it is possible to assess their impact in aquatic systems. A review is also provided on the status and likely value of current methods used in monitoring and impact assessment. Information is included on environmental chemistry, the pollution-ecology of algae, invertebrates and fish, and on aquatic toxicology, genetic toxicology, and the pathology of fishes and invertebrates in relation to heavy metals.

  10. Local stability tests in Dresden 2 boiling water reactor

    SciTech Connect

    March-Leuba, J.; Fry, D.N.; Buchanan, M.E.; McNew, C.O.

    1984-04-01

    This report presents the results of a local stability test performed at Dresden Unit 2 in May 1983 to determine the effect of a new fuel element design on local channel stability. This test was performed because the diameter of the new fuel rods increases the heat transfer coefficient, making the reactor more responsive and, thus, more susceptible to instabilities. After four of the new fuel elements with a 9 x 9 array of fuel rods were loaded into Dresden 2, the test was performed by inserting an adjacent control rod all the way in and then withdrawing it to its original position at maximum speed. At the moment of the test, reactor conditions were 52.7% power and 38.9% flow. Both the new 9 x 9 fuel elements and the standard 8 x 8 ones proved to be locally stable when operating at minimum pump speed at the beginning of cycle in Dresden 2, and no significant difference was found between the behavior of the two fuel types. Finally, Dresden 2 showed a high degree of stability during control rod and normal noise type perturbations.

  11. LIGHT WATER REACTOR ACCIDENT TOLERANT FUELS IRRADIATION TESTING

    SciTech Connect

    Carmack, William Jonathan; Barrett, Kristine Eloise; Chichester, Heather Jean MacLean

    2015-09-01

    The purpose of Accident Tolerant Fuels (ATF) experiments is to test novel fuel and cladding concepts designed to replace the current zirconium alloy uranium dioxide (UO2) fuel system. The objective of this Research and Development (R&D) is to develop novel ATF concepts that will be able to withstand loss of active cooling in the reactor core for a considerably longer time period than the current fuel system while maintaining or improving the fuel performance during normal operations, operational transients, design basis, and beyond design basis events. It was necessary to design, analyze, and fabricate drop-in capsules to meet the requirements for testing under prototypic LWR temperatures in Idaho National Laboratory's Advanced Test Reactor (ATR). Three industry led teams and one DOE team from Oak Ridge National Laboratory provided fuel rodlet samples for their new concepts for ATR insertion in 2015. As-built projected temperature calculations were performed on the ATF capsules using the BISON fuel performance code. BISON is an application of INL’s Multi-physics Object Oriented Simulation Environment (MOOSE), which is a massively parallel finite element based framework used to solve systems of fully coupled nonlinear partial differential equations. Both 2D and 3D models were set up to examine cladding and fuel performance.

  12. Spatial Burnout in Water Reactors with Nonuniform Startup Distributions of Uranium and Boron

    NASA Technical Reports Server (NTRS)

    Fox, Thomas A.; Bogart, Donald

    1955-01-01

    Spatial burnout calculations have been made of two types of water moderated cylindrical reactor using boron as a burnable poison to increase reactor life. Specific reactors studied were a version of the Submarine Advanced Reactor (sAR) and a supercritical water reactor (SCW) . Burnout characteristics such as reactivity excursion, neutron-flux and heat-generation distributions, and uranium and boron distributions have been determined for core lives corresponding to a burnup of approximately 7 kilograms of fully enriched uranium. All reactivity calculations have been based on the actual nonuniform distribution of absorbers existing during intervals of core life. Spatial burnout of uranium and boron and spatial build-up of fission products and equilibrium xenon have been- considered. Calculations were performed on the NACA nuclear reactor simulator using two-group diff'usion theory. The following reactor burnout characteristics have been demonstrated: 1. A significantly lower excursion in reactivity during core life may be obtained by nonuniform rather than uniform startup distribution of uranium. Results for SCW with uranium distributed to provide constant radial heat generation and a core life corresponding to a uranium burnup of 7 kilograms indicated a maximum excursion in reactivity of 2.5 percent. This compared to a maximum excursion of 4.2 percent obtained for the same core life when w'anium was uniformly distributed at startup. Boron was incorporated uniformly in these cores at startup. 2. It is possible to approach constant radial heat generation during the life of a cylindrical core by means of startup nonuniform radial and axial distributions of uranium and boron. Results for SCW with nonuniform radial distribution of uranium to provide constant radial heat generation at startup and with boron for longevity indicate relatively small departures from the initially constant radial heat generation distribution during core life. Results for SAR with a sinusoidal

  13. Reconstructing the direction of reactor antineutrinos via electron scattering in Gd-doped water Cherenkov detectors

    NASA Astrophysics Data System (ADS)

    Hellfeld, D.; Bernstein, A.; Dazeley, S.; Marianno, C.

    2017-01-01

    The potential of elastic antineutrino-electron scattering in a Gd-doped water Cherenkov detector to determine the direction of a nuclear reactor antineutrino flux was investigated using the recently proposed WATCHMAN antineutrino experiment as a baseline model. The expected scattering rate was determined assuming a 13-km standoff from a 3.758-GWt light water nuclear reactor and the detector response was modeled using a Geant4-based simulation package. Background was estimated via independent simulations and by scaling published measurements from similar detectors. Background contributions were estimated for solar neutrinos, misidentified reactor-based inverse beta decay interactions, cosmogenic radionuclides, water-borne radon, and gamma rays from the photomultiplier tubes (PMTs), detector walls, and surrounding rock. We show that with the use of low background PMTs and sufficient fiducialization, water-borne radon and cosmogenic radionuclides pose the largest threats to sensitivity. Directional sensitivity was then analyzed as a function of radon contamination, detector depth, and detector size. The results provide a list of experimental conditions that, if satisfied in practice, would enable antineutrino directional reconstruction at 3σ significance in large Gd-doped water Cherenkov detectors with greater than 10-km standoff from a nuclear reactor.

  14. Reconstructing the direction of reactor antineutrinos via electron scattering in Gd-doped water Cherenkov detectors

    DOE PAGES

    Hellfeld, D.; Bernstein, A.; Dazeley, S.; ...

    2016-10-17

    The potential of elastic antineutrino-electron scattering in a Gd-doped water Cherenkov detector to determine the direction of a nuclear reactor antineutrino flux was investigated using the recently proposed WATCHMAN antineutrino experiment as a baseline model. The expected scattering rate was determined assuming a 13 km standoff from a 3.758 GWt light water nuclear reactor and the detector response was modeled using a Geant4-based simulation package. Background was estimated via independent simulations and by scaling published measurements from similar detectors. Background contributions were estimated for solar neutrinos, misidentified reactor-based inverse beta decay interactions, cosmogenic radionuclides, water-borne radon, and gamma rays frommore » the photomultiplier tubes, detector walls, and surrounding rock. We show that with the use of low background PMTs and sufficient fiducialization, water-borne radon and cosmogenic radionuclides pose the largest threats to sensitivity. Directional sensitivity was then analyzed as a function of radon contamination, detector depth, and detector size. Lastly, the results provide a list of experimental conditions that, if satisfied in practice, would enable antineutrino directional reconstruction at 3 sigma significance in large Gd-doped water Cherenkov detectors with greater than 10 km standoff from a nuclear reactor.« less

  15. Environmentally assisted cracking in light water reactors. Semiannual progress report, January 1996--June 1996

    SciTech Connect

    Chopra, O.K.; Chung, H.M.; Gruber, E.E.

    1997-05-01

    This report summarizes work performed by Argonne National Laboratory on fatigue and environmentally assisted cracking (EAC) in light water reactors from January 1996 to June 1996. Topics that have been investigated include (a) fatigue of carbon, low-alloy, and austenitic stainless steels (SSs) used in reactor piping and pressure vessels, (b) irradiation-assisted stress corrosion cracking of Type 304 SS, and (c) EAC of Alloys 600 and 690. Fatigue tests were conducted on ferritic and austenitic SSs in water that contained various concentrations of dissolved oxygen (DO) to determine whether a slow strain rate applied during various portions of a tensile-loading cycle are equally effective in decreasing fatigue life. Slow-strain-rate-tensile tests were conducted in simulated boiling water reactor (BWR) water at 288{degrees}C on SS specimens irradiated to a low fluence in the Halden reactor and the results were compared with similar data from a control-blade sheath and neutron-absorber tubes irradiated in BWRs to the same fluence level. Crack-growth-rate tests were conducted on compact-tension specimens from several heats of Alloys 600 and 690 in air and high-purity, low-DO water. 83 refs., 60 figs., 14 tabs.

  16. Environmentally assisted cracking in light water reactors. Semiannual report July 1996--December 1996

    SciTech Connect

    Chopra, O.K.; Chung, H.M.; Gavenda, D.J.

    1997-10-01

    This report summarizes work performed by Argonne National Laboratory on fatigue and environmentally assisted cracking (EAC) in light water reactors from July 1996 to December 1996. Topics that have been investigated include (a) fatigue of carbon, low-alloy, and austenitic stainless steels (SSs) used in reactor piping and pressure vessels, (b) irradiation-assisted stress corrosion cracking of Type 304 SS, (c) EAC of Alloy 600, and (d) characterization of residual stresses in welds of boiling water reactor (BWR) core shrouds by numerical models. Fatigue tests were conducted on ferritic and austenitic SSs in water that contained various concentrations of dissolved oxygen to determine whether a slow strain rate applied during various portions of a tensile-loading cycle are equally effective in decreasing fatigue life. Slow-strain-rate-tensile tests were conducted in simulated BWR water at 288 C on SS specimens irradiated to a low fluence in the Halden reactor and the results were compared with similar data from a control-blade sheath and neutron-absorber tubes irradiated in BWRs to the same fluence level. Crack-growth-rate tests were conducted on compact-tension specimens from a low-carbon content heat of Alloy 600 in high-purity oxygenated water at 289 C. Residual stresses and stress intensity factors were calculated for BWR core shroud welds.

  17. Reconstructing the direction of reactor antineutrinos via electron scattering in Gd-doped water Cherenkov detectors

    SciTech Connect

    Hellfeld, D.; Bernstein, A.; Dazeley, S.; Marianno, C.

    2016-10-17

    The potential of elastic antineutrino-electron scattering in a Gd-doped water Cherenkov detector to determine the direction of a nuclear reactor antineutrino flux was investigated using the recently proposed WATCHMAN antineutrino experiment as a baseline model. The expected scattering rate was determined assuming a 13 km standoff from a 3.758 GWt light water nuclear reactor and the detector response was modeled using a Geant4-based simulation package. Background was estimated via independent simulations and by scaling published measurements from similar detectors. Background contributions were estimated for solar neutrinos, misidentified reactor-based inverse beta decay interactions, cosmogenic radionuclides, water-borne radon, and gamma rays from the photomultiplier tubes, detector walls, and surrounding rock. We show that with the use of low background PMTs and sufficient fiducialization, water-borne radon and cosmogenic radionuclides pose the largest threats to sensitivity. Directional sensitivity was then analyzed as a function of radon contamination, detector depth, and detector size. Lastly, the results provide a list of experimental conditions that, if satisfied in practice, would enable antineutrino directional reconstruction at 3 sigma significance in large Gd-doped water Cherenkov detectors with greater than 10 km standoff from a nuclear reactor.

  18. Environmentally assisted cracking in Light Water Reactors: Semiannual report, April 1993--September 1993. Volume 17

    SciTech Connect

    Chopra, O.K.; Chung, H.M.; Karlsen, T.; Kassner, T.F.; Michaud, W.F.; Ruther, W.E.; Sanecki, J.E.; Shack, W.J.; Soppet, W.K.

    1994-06-01

    This report summarizes work performed by Argonne National Laboratory on fatigue and environmentally assisted cracking (EAC) in light water reactors (LWRS) during the six months from April 1993 to September 1993. EAC and fatigue of piping, pressure vessels, and core components in LWRs are important concerns as extended reactor lifetimes are envisaged. Topics that have been investigated include (a) fatigue of low-alloy steel used in piping, steam generators, and reactor pressure vessels; (b) EAC of cast stainless steels (SSs); and (c) radiation-induced segregation and irradiation-assisted stress corrosion cracking of Type 304 SS after accumulation of relatively high fluence. Fatigue tests were conducted on medium-sulfur-content A106-Gr B piping and A533-Gr B pressure vessel steels in simulated PWR water and in air. Additional crack growth data were obtained on fracture-mechanics specimens of cast austenitic SSs in the as-received and thermally aged conditions in simulated boiling-water reactor (BWR) water at 289{degree}C. The data were compared with predictions based on crack growth correlations for wrought austenitic SS in oxygenated water developed at ANL and rates in air from Section 11 of the ASME Code. Microchemical and microstructural changes in high- and commercial-purity Type 304 SS specimens from control-blade absorber tubes and a control-blade sheath from operating BWRs were studied by Auger electron spectroscopy and scanning electron microscopy.

  19. Environmentally assisted cracking in light water reactors. Semiannual report, April 1994--September 1994, Volume 19

    SciTech Connect

    Chopra, O.K.; Chung, H.M.; Gavenda, D.J.

    1995-09-01

    This report summarizes work performed by Argonne National Laboratory (ANL) on fatigue and environmentally assisted cracking (EAC) in light water reactors from April to September 1994. Topics that have been investigated include (a) fatigue of carbon and low-alloy steel used in piping and reactor pressure vessels, (b) EAC of austenitic stainless steels (SSs) and Alloy 600, and (c) irradiation-assisted stress corrosion cracking (IASCC) of Type 304 SS. Fatigue tests have been conducted on A106-Gr B and A533-Gr B steels in oxygenated water to determine whether a slow strain rate applied during different portions of a tensile-loading cycle are equally effective in decreasing fatigue life. Crack growth data were obtained on fracture-mechanics specimens of SSs and Alloy 600 to investigate EAC in simulated boiling water reactor (BWR) and pressurized water reactor environments at 289{degrees}C. The data were compared with predictions from crack growth correlations developed at ANL for SSs in water and from rates in air from Section XI of the ASME Code. Microchemical changes in high- and commercial-purity Type 304 SS specimens from control-blade absorber tubes and a control-blade sheath from operating BWRs were studied by Auger electron spectroscopy and scanning electron microscopy to determine whether trace impurity elements may contribute to IASCC of these materials.

  20. Implementation of a source term control program in a mature boiling water reactor.

    PubMed

    Vargo, G J; Jarvis, A J; Remark, J F

    1991-06-01

    The implementation and results of a source term control program implemented at the James A. FitzPatrick Nuclear Power Plant (JAF), a mature boiling water reactor (BWR) facility that has been in commercial operation since 1975, are discussed. Following a chemical decontamination of the reactor water recirculation piping in the Reload 8/Cycle 9 refueling outage in 1988, hydrogen water chemistry (HWC) and feedwater Zn addition were implemented. This is the first application of both HWC and feedwater Zn addition in a BWR facility. The radiological benefits and impacts of combined operation of HWC and feedwater Zn addition at JAF during Cycle 9 are detailed and summarized. The implementation of hydrogen water chemistry resulted in a significant transport of corrosion products within the reactor coolant system that was greater than anticipated. Feedwater Zn addition appears to be effective in controlling buildup of other activated corrosion products such as 60Co on reactor water recirculation piping; however, adverse impacts were encountered. The major adverse impact of feedwater Zn addition is the production of 65Zn that is released during plant outages and operational transients.

  1. Spatial distribution and multiple sources of heavy metals in the water of Chaohu Lake, Anhui, China.

    PubMed

    Li, Guolian; Liu, Guijian; Zhou, Chuncai; Chou, Chen-Lin; Zheng, Liugen; Wang, Jizhong

    2012-05-01

    In this study, a survey for the spatial distribution of heavy metals in Chaohu Lake of China was conducted. Sixty-two surface water samples were collected from entire lake including three of its main river entrances. This is the first systematic report concerning the content, distribution, and origin of heavy metals (Cu, Cr, Cd, Hg, Zn, and Ni) in the Chaohu Lake water. The results showed that heavy metals (Cu, Cr, Zn, and Ni) concentrations in the estuary of Nanfei River were relatively higher than those in the other areas, while content of Hg is higher in the southeast lake than northwest lake. Moreover, Cd has locally concentration in the surface water from the entire Chaohu Lake. The heavy metal average concentrations, except Hg, were lower than the cutoff values for the first-grade water quality (China Environment Quality Standard) which was set as the highest standard to protect the social nature reserves. The Hg content is between the grades three and four water quality, and other heavy metals contents are higher than background values. The aquatic environment of Chaohu Lake has apparently been contaminated. Both the cluster analysis (CA) and correlation analysis provide information about the origin of heavy metals in the Lake. Our findings indicated that agricultural activities and adjacent plants chimneys may contribute the most to Cd and Hg contamination of Chaohu Lake, respectively.

  2. Flaw density examinations of a clad boiling water reactor pressure vessel segment

    SciTech Connect

    Cook, K.V.; McClung, R.W.

    1986-01-01

    Flaw density is the greatest uncertainty involved in probabilistic analyses of reactor pressure vessel failure. As part of the Heavy-Section Steel Technology (HSST) Program, studies have been conducted to determine flaw density in a section of reactor pressure vessel cut from the Hope Creek Unit 2 vessel (nominally 0.7 by 3 m (2 by 10 ft)). This section (removed from the scrapped vessel that was never in service) was evaluated nondestructively to determine the as-fabricated status. We had four primary objectives: (1) evaluate longitudinal and girth welds for flaws with manual ultrasonics, (2) evaluate the zone under the nominal 6.3-mm (0.25-in.) clad for cracking (again with manual ultrasonics), (3) evaluate the cladding for cracks with a high-sensitivity fluorescent penetrant method, and (4) determine the source of indications detected.

  3. Stacked waveguide reactors with gradient embedded scatterers for high-capacity water cleaning

    DOE PAGES

    Ahsan, Syed Saad; Gumus, Abdurrahman; Erickson, David

    2015-11-04

    We present a compact water-cleaning reactor with stacked layers of waveguides containing gradient patterns of optical scatterers that enable uniform light distribution and augmented water-cleaning rates. Previous photocatalytic reactors using immersion, external, or distributive lamps suffer from poor light distribution that impedes scalability. Here, we use an external UV-source to direct photons into stacked waveguide reactors where we scatter the photons uniformly over the length of the waveguide to thin films of TiO2-catalysts. In conclusion, we also show 4.5 times improvement in activity over uniform scatterer designs, demonstrate a degradation of 67% of the organic dye, and characterize the degradationmore » rate constant.« less

  4. Nanostructure of Metallic Particles in Light Water Reactor Used Nuclear Fuel

    SciTech Connect

    Buck, Edgar C.; Mausolf, Edward J.; Mcnamara, Bruce K.; Soderquist, Chuck Z.; Schwantes, Jon M.

    2015-03-11

    The extraordinary nano-structure of metallic particles in light water reactor fuels points to possible high reactivity through increased surface area and a high concentration of high energy defect sites. We have analyzed the metallic epsilon particles from a high burn-up fuel from a boiling water reactor using transmission electron microscopy and have observed a much finer nanostructure in these particles than has been reported previously. The individual round particles that varying in size between ~20 and ~50 nm appear to consist of individual crystallites on the order of 2-3 nm in diameter. It is likely that in-reactor irradiation induce displacement cascades results in the formation of the nano-structure. The composition of these metallic phases is variable yet the structure of the material is consistent with the hexagonal close packed structure of epsilon-ruthenium. These findings suggest that unusual catalytic behavior of these materials might be expected, particularly under accident conditions.

  5. Stacked waveguide reactors with gradient embedded scatterers for high-capacity water cleaning

    SciTech Connect

    Ahsan, Syed Saad; Gumus, Abdurrahman; Erickson, David

    2015-11-04

    We present a compact water-cleaning reactor with stacked layers of waveguides containing gradient patterns of optical scatterers that enable uniform light distribution and augmented water-cleaning rates. Previous photocatalytic reactors using immersion, external, or distributive lamps suffer from poor light distribution that impedes scalability. Here, we use an external UV-source to direct photons into stacked waveguide reactors where we scatter the photons uniformly over the length of the waveguide to thin films of TiO2-catalysts. In conclusion, we also show 4.5 times improvement in activity over uniform scatterer designs, demonstrate a degradation of 67% of the organic dye, and characterize the degradation rate constant.

  6. Qualification of a Tritium-Producing Target for the light water reactor application

    SciTech Connect

    Apley, W.J.; Beeman, G.H.; Ethridge, J.L.

    1992-06-01

    The Pacific Northwest Laboratory (PNL) currently manages the Light Water Reactor (LWR) Tritium Target Development Project (TTDP) for the Office of New Production Reactors (NP), US Department of Energy. The project`s objective is to demonstrate and qualify a high-temperature LWR tritium target system with fabrication and extraction processes sufficiently confirmed to ensure a deployable system consistent with variable tritium production demands. The project has also examined and reported on technical and institutional issues associated with acquisition and conversion of the yet uncompleted Washington Public Power Supply System Unit I (WNP-1) for tritium production purposes. WNP-1 is a 63% complete, Babcock and Wilcox, 3800 MW thermal, 205 assembly, pressurized water reactor located at Hanford, Washington. WNP-1 has served as the reference LWR plant for the technical evaluation and target development activities. This report discussed the evaluation and development necessary to provide a complete LWR target qualification package.

  7. TRAC-PF1: an advanced best-estimate computer program for pressurized water reactor analysis

    SciTech Connect

    Liles, D.R.; Mahaffy, J.H.

    1984-02-01

    The Transient Reactor Analysis Code (TRAC) is being developed at the Los Alamos National Laboratory to provide advanced best-estimate predictions of postulated accidents in light water reactors. The TRAC-PF1 program provides this capability for pressurized water reactors and for many thermal-hydraulic experimental facilities. The code features either a one-dimensional or a three-dimensional treatment of the pressure vessel and its associated internals; a two-phase, two-fluid nonequilibrium hydrodynamics model with a noncondensable gas field; flow-regime-dependent constitutive equation treatment; optional reflood tracking capability for both bottom flood and falling-film quench fronts; and consistent treatment of entire accident sequences including the generation of consistent initial conditions. This report describes the thermal-hydraulic models and the numerical solution methods used in the code. Detailed programming and user information also are provided.

  8. Technical specification: Mixed-oxide pellets for the light-water reactor irradiation demonstration test

    SciTech Connect

    Cowell, B.S.

    1997-06-01

    This technical specification is a Level 2 Document as defined in the Fissile Materials Disposition Program Light-Water Reactor Mixed-oxide Fuel Irradiation Test Project Plan. It is patterned after the pellet specification that was prepared by Atomic Energy of Canada, Limited, for use by Los Alamos National Laboratory in fabrication of the test fuel for the Parallex Project, adjusted as necessary to reflect the differences between the Canadian uranium-deuterium reactor and light-water reactor fuels. This specification and the associated engineering drawing are to be utilized only for preparation of test fuel as outlined in the accompanying Request for Quotation and for additional testing as directed by Oak Ridge National Laboratory or the Department of Energy.

  9. USING BIOPOLYMERS TO REMOVE HEAVY METALS FROM SOIL AND WATER

    EPA Science Inventory

    Chemical remediation of soil may involve the use of harsh chemicals that generate waste streams, which may adversely affect the soil's integrity and ability to support vegetation. This article reviews the potential use of benign reagents, such as biopolymers, to extract heavy me...

  10. USING BIOPOLYMERS TO REMOVE HEAVY METALS FROM SOIL AND WATER

    EPA Science Inventory

    Chemical remediation of soil may involve the use of harsh chemicals that generate waste streams, which may adversely affect the soil's integrity and ability to support vegetation. This article reviews the potential use of benign reagents, such as biopolymers, to extract heavy me...

  11. Using biopolymers to remove heavy metals from soil and water

    SciTech Connect

    Krishnamurthy, S.; Frederick, R.M.

    1993-11-19

    Chemical remediation of soil may involve the use of harsh chemicals that generate waste streams, which may adversely affect the soil's integrity and ability to support vegetation. This article reviews the potential use of benign reagents, such as biopolymers, to extract heavy metals. The biopolymers discussed are chitin and chitosan, modified starch, cellulose, and polymer-containing algae. (Copyright (c) Remediation 1994.)

  12. EXPERIMENTAL EVALUATION OF THE THERMAL PERFORMANCE OF A WATER SHIELD FOR A SURFACE POWER REACTOR

    SciTech Connect

    REID, ROBERT S.; PEARSON, J. BOSIE; STEWART, ERIC T.

    2007-01-16

    Water based reactor shielding is being investigated for use on initial lunar surface power systems. A water shield may lower overall cost (as compared to development cost for other materials) and simplify operations in the setup and handling. The thermal hydraulic performance of the shield is of significant interest. The mechanism for transferring heat through the shield is natural convection. Natural convection in a 100 kWt lunar surface reactor shield design is evaluated with 2 kW power input to the water in the Water Shield Testbed (WST) at the NASA Marshall Space Flight Center. The experimental data from the WST is used to validate a CFD model. Performance of the water shield on the lunar surface is then predicted with a CFD model anchored to test data. The experiment had a maximum water temperature of 75 C. The CFD model with 1/6-g predicts a maximum water temperature of 88 C with the same heat load and external boundary conditions. This difference in maximum temperature does not greatly affect the structural design of the shield, and demonstrates that it may be possible to use water for a lunar reactor shield.

  13. Experimental Evaluation of the Thermal Performance of a Water Shield for a Surface Power Reactor

    SciTech Connect

    Pearson, J. Boise; Stewart, Eric T.; Reid, Robert S.

    2007-01-30

    Water based reactor shielding is being investigated for use on initial lunar surface power systems. A water shield may lower overall cost (as compared to development cost for other materials) and simplify operations in the setup and handling. The thermal hydraulic performance of the shield is of significant interest. The mechanism for transferring heat through the shield is natural convection. Natural convection in a 100 kWt lunar surface reactor shield design is evaluated with 2 kW power input to the water in the Water Shield Testbed (WST) at the NASA Marshall Space Flight Center. The experimental data from the WST is used to validate a CFD model. Performance of the water shield on the lunar surface is then predicted with a CFD model anchored to test data. The experiment had a maximum water temperature of 75 deg. C. The CFD model with 1/6-g predicts a maximum water temperature of 88 deg. C with the same heat load and external boundary conditions. This difference in maximum temperature does not greatly affect the structural design of the shield, and demonstrates that it may be possible to use water for a lunar reactor shield.

  14. Chemometric evaluation of the heavy metals distribution in waters from the Dilovasi region in Kocaeli, Turkey.

    PubMed

    Bingöl, Deniz; Ay, Umit; Karayünlü Bozbaş, Seda; Uzgören, Nevin

    2013-03-15

    The main objective of this study was to test water samples collected from 10 locations in the Dilovası area (a town in the Kocaeli region of Turkey) for heavy metal contamination and to classify the heavy metal (Cr, Mn, Co, Ni, Cu, Zn, As, Cd, Pb and Hg) contents in water samples using chemometric methods. The heavy metals in the water samples were identified using inductively coupled plasma-mass spectrometry (ICP-MS). To ascertain the relationship among the water samples and their possible sources, the correlation analysis, principal component analysis (PCA), and cluster analysis (CA) were used as classification techniques. About 10 water samples were classified into five groups using PCA. A very similar grouping was obtained using CA.

  15. Assessments of Water Ingress Accidents in a Modular High-Temperature Gas-Cooled Reactor

    SciTech Connect

    Zhang Zuoyi; Dong Yujie; Scherer, Winfried

    2005-03-15

    Severe water ingress accidents in the 200-MW HTR-module were assessed to determine the safety margins of modular pebble-bed high-temperature gas-cooled reactors (HTR-module). The 200-MW HTR-module was designed by Siemens under the criteria that no active safety protection systems were necessary because of its inherent safe nature. For simulating the behavior of the HTR-module during severe water ingress accidents, a water, steam, and helium multiphase cavity model was developed and implemented in the dynamic simulator for nuclear power plants (DSNP) simulation system. Comparisons of the DSNP simulations incorporating these models with experiments and with calculations using the time-dependent neutronics and temperature dynamics code were made to validate the simulation. The analysis of the primary circuit showed that the maximum water concentration increase in the reactor core was <0.3 kg/(m{sup 3}s). The water vaporization in the steam generator and characteristics of water transport from the steam generator to the reactor core would reduce the rate of water ingress into the reactor core. The analysis of a full cavitation of the feedwater pump showed that if the secondary circuit could be depressurized, the feedwater pump would be stopped by the full cavitation. This limits the water transported from the deaerator to the steam generator. A comprehensive simulation of the HTR-module power plant showed that the water inventory in the primary circuit was limited to {approx}3000 kg. The nuclear reactivity increase caused by the water ingress would lead to a fast power excursion, which would be inherently counterbalanced by negative feedback effects. The integrity of the fuel elements, because the safety-relevant temperature limit of 1600 deg. C is not reached in any case, is not challenged.

  16. Hydraulic performance of a proposed in situ photocatalytic reactor for degradation of MTBE in water.

    PubMed

    Lim, Leonard Lik Pueh; Lynch, Rod

    2011-01-01

    Methyl tert-butyl ether (MTBE) groundwater remediation projects often require a combination of technologies resulting in increasing the project costs. A cost-effective in situ photocatalytic reactor design, Honeycomb II, is proposed and tested for its efficiency in MTBE degradation at various flows. This study is an intermediate phase of the research in developing an in situ photocatalytic reactor for groundwater remediation. It examines the effect of the operating variables: air and water flow and double passages through Honeycomb II, on the MTBE removal. MTBE vaporisation is affected by not only temperature, Henry's law constant and air flow to volume ratio but also reactor geometry. The column reactor achieved more than 84% MTBE removal after 8 h at flows equivalent to horizontal groundwater velocities slower than 21.2 cm d⁻¹. Despite the contrasting properties between a photocatalytic indicator methylene blue and MTBE, the reactor efficiency in degrading both compounds showed similar responses towards flow (equivalent groundwater velocity and hydraulic residence time (HRT)). The critical HRT for both compounds was approximately 1 d, which corresponded to a velocity of 21.2 cm d⁻¹. A double pass through both new and used catalysts achieved more than 95% MTBE removal after two passes in 48 h. It also verified that the removal efficiency can be estimated via the sequential order of the removal efficiency of one pass obtained in the laboratory. This study reinforces the potential of this reactor design for in situ groundwater remediation.

  17. The simulation of thermohydraulic phenomena in a pressurized water reactor primary loop

    SciTech Connect

    Popp, M

    1987-01-01

    Several important fluid flow and heat transfer phenomena essential to nuclear power reactor safety were investigated. Scaling and modeling laws for pressurized water reactors are reviewed and a new scaling approach focusing on the overall loop behavior is presented. Scaling criteria for one- and two-phase natural circulation are developed, as well as a simplified model describing the first phase of a small break loss of coolant accident. Reactor vessel vent valve effects are included in the analysis of steady one-phase natural circulation flow. Two new dimensionless numbers, which uniquely describe one-phase flow in natural circulation loops, were deduced and are discussed. A scaled model of the primary loop of a typical Babcock and Wilcox reactor was designed, built, and tested. The particular prototype modeled was the TMI unit 2 reactor. The electrically heated, stainless steel model operates at a maximum pressure of 300 psig and has a maximum heat input of 188 kW. The model is about 4 times smaller in height than the prototype reactor, with a nominal volume scale of 1:500. Experiments were conducted establishing subcooled natural circulation in the model loop. Both steady flow and power transients were investigated.

  18. Heavy metal release from different ashes during serial batch tests using water and acid.

    PubMed

    Ludwig, Bernard; Khanna, Partap; Prenzel, Jürgen; Beese, Friedrich

    2005-01-01

    Most ashes contain a significant amount of heavy metals and when released from disposed or used ash materials, they can form a major environmental concern for underground waters. The use of water extracts to assess the easily mobilisable content of heavy metals may not provide an appropriate measure. This study describes the patterns of heavy metal release from ash materials in context with results from the German standard extraction method DIN-S4 (DIN 38 414 S4). Samples of four different ashes (municipal solid waste incineration ash, wood ash, brown coal ash and hard coal ash) were subjected to a number of serial batch tests with liquid renewal, some of which involved the addition of acid to neutralize carbonates and oxides. Release of heavy metals showed different patterns depending on the element, the type of material, the method of extraction and the type of the extractant used. Only a small fraction of the total heavy metal contents occurred as water soluble salts; of special significance was the amount of Cr released from the wood ash. The reaction time (1, 24 or 72 h between each extraction step with water) had only a small effect on the release of heavy metals. However, the release of most of the heavy metals was governed by the dissolution processes following proton inputs, indicating that pH-dependent tests such as CEN TC 292 or others are required to estimate long-term effects of heavy metal releases from ashes. Based on the chemical characteristics of ash materials in terms of their form and solubility of heavy metals, recommendations were made on the disposal or use of the four ash materials.

  19. Environmentally assisted cracking in light water reactors. Semiannual report, October 1993--March 1994. Volume 18

    SciTech Connect

    Chung, H.M.; Chopra, O.K.; Erck, R.A.; Kassner, T.F.; Michaud, W.F.; Ruther, W.E.; Sanecki, J.E.; Shack, W.J.; Soppet, W.K.

    1995-03-01

    This report summarizes work performed by Argonne National Laboratory (ANL) on fatigue and environmentally assisted cracking (EAC) in light water reactors (LWRs) during the six months from October 1993 to March 1994. EAC and fatigue of piping, pressure vessels, and core components in LWRs are important concerns in operating plants and as extended reactor lifetimes are envisaged. Topics that have been investigated include (a) fatigue of low-alloy steel used in piping, steam generators, and reactor pressure vessels, (b) EAC of wrought and cast austenitic stainless steels (SSs), and (c) radiation-induced segregation and irradiation-assisted stress corrosion cracking (IASCC) of Type 304 SS after accumulation of relatively high fluence. Fatigue tests have been conducted on A302-Gr B low-alloy steel to verify whether the current predictions of modest decreases of fatigue life in simulated pressurized water reactor water are valid for high-sulfur heats that show environmentally enhanced fatigue crack growth rates. Additional crack growth data were obtained on fracture-mechanics specimens of austenitic SSs to investigate threshold stress intensity factors for EAC in high-purity oxygenated water at 289{degrees}C. The data were compared with predictions based on crack growth correlations for wrought austenitic SS in oxygenated water developed at ANL and rates in air from Section XI of the ASME Code. Microchemical and microstructural changes in high- and commercial-purity Type 304 SS specimens from control-blade absorber tubes and a control-blade sheath from operating boiling water reactors were studied by Auger electron spectroscopy and scanning electron microscopy to determine whether trace impurity elements, which are not specified in the ASTM specifications, may contribute to IASCC of solution-annealed materials.

  20. Heavy metal distribution and water quality characterization of water bodies in Louisiana's Lake Pontchartrain Basin, USA.

    PubMed

    Zhang, Zengqiang; Wang, Jim J; Ali, Amjad; DeLaune, Ronald D

    2016-11-01

    The seasonal variation in physico-chemical properties, anions, and the heavy metal (Cd, Co, Cr, Cu, Mn, Ni, Pb, and Zn) concentration was evaluated in water from nine different rivers in Lake Pontchartrain Basin, Louisiana, USA. The water quality parameters were compared with toxicity reference values (TRV), US Environmental Protection Agency (USEPA) drinking/aquatic life protection, and WHO standards. Among physico-chemical properties, pH, DO, and turbidity were high during spring, while, EC, temperature, and DOC were high during summer and vice versa. The anion study revealed that the concentrations of F(-), Cl(-), and NO3(-) were higher during summer and Br(-) and SO4(-) were higher during spring. Our research findings showed anion concentration decreased in the order of Cl(-) > SO4(-) > NO3(-) > Br(-) > F(-), in accordance with the global mean anion concentration. The dissolved heavy metals (Cd, Co, Cr, Cu, Mn, Ni, Pb) except Zn were higher during spring than summer. None of the rivers showed any Cd pollution for both seasons. Co showed higher concentrations in Amite River, Mississippi River, Industrial Canal, and Lacombe Bayou during summer. The Cr concentration was higher than WHO drinking water standards, implicating water unsuitability for drinking purposes in all the rivers associated with the Lake Pontchartrain Basin. Cu showed no pollution risk for the study area. Mn and Co were similar to concentration in Lacombe Bayou, Liberty Bayou, Blind River, and Industrial Canal. Mn levels were greater than WHO standards for the Tickfaw River, Tangipahoa River, and Blind River in both seasons. Blind River, Tangipahoa River, Tickfaw River, and Amite River will require more monitoring for determining possible Mn pollution. Ni content in river water during both seasons showed low pollution risk. Liberty Bayou and Industrial Canal concentrations were closer to the WHO regulatory standards, indicating possible risk of Pb pollution in these water bodies. The Zn

  1. TMI-2 Reactor Building source term measurements: surfaces and basement water and sediment

    SciTech Connect

    McIsaac, C V; Keefer, D G

    1984-10-01

    Presented in this report are the results of radiochemical and elemental analyses performed on samples collected from the Three Mile Island Unit 2 Reactor Building from August 1979 to December 1983. The quantities of fission products and core materials that were measured on the external surfaces in the Reactor Building or in the water and sediment in its basement are summarized. Recent analysis results for access panels removed from the air cooling assembly and for liquid and particulate samples collected from the Reactor Building sump and reactor coolant drain tank are included in the report. Measurements show that 59% of the /sup 3/H, 2.7% of the /sup 90/Sr, 15% of the /sup 129/I, 20% of the /sup 131/I, and 42% of the /sup 137/Cs originally in the core at the time of the accident could be accounted for outside the core in the Reactor Building. With the exceptions of /sup 90/Sr and /sup 144/Ce, the vast majority of each radionuclide released was found dispersed in the water and sediment in the basement.

  2. Nuclear Engineering Computer Modules, Thermal-Hydraulics, TH-1: Pressurized Water Reactors.

    ERIC Educational Resources Information Center

    Reihman, Thomas C.

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

  3. Light water reactor safety research program. Volume 12: quarterly report, Apr-Jun 79

    SciTech Connect

    Berman, M.

    1980-05-01

    This report summarizes the progress of the Light Water Reactor Safety Research Program during the 2nd quarter of 1979. Specifically, the report summarizes progress in five major areas of research. They are: (1) the molten core/concrete interactions study; (2) steam explosion research phenomena; (3) statistical LOCA analysis; (4) UHI model development; (5) two-phase jet loads.

  4. A computer program for estimating decommissioning costs for light water reactors

    SciTech Connect

    Bierschbach, M.C.

    1993-02-01

    This report discusses a desk-top computer program has been developed for estimating the costs, waste volumes, and occupational radiation exposures associated with decommissioning light-water reactor power stations. Cost categories and cost algorithms used in the program are discussed and a brief description of the user interface is given.

  5. Environmental degradation of materials in nuclear power systems-water reactors

    SciTech Connect

    Not Available

    1984-01-01

    This book presents the papers given at a conference which focused on corrosion damage in light water reactors. Topics considered include material degradation issues in key components, material degradation in service, microstructural and compositional effects, the effects of the environment, the effects of mechanical variables, and environment and material remedies.

  6. Environmentally assisted cracking in light water reactors. Semiannual report, April--September 1991: Volume 13

    SciTech Connect

    Kassner, T F; Ruther, W E; Chung, H M; Hicks, P D; Hins, A G; Park, J Y; Soppet, W K; Shack, W J

    1992-03-01

    This report summarizes work performed by Argonne National Laboratory on fatigue and environmentally assisted cracking in high water reactors during the six months from April 1991 through September 1991. Topics that have been investigated during this period include (1) fatigue and stress corrosion cracking (SCC) of low-alloy steel used in piping and in steam generator and reactor pressure vessels; (2) role of chromate and sulfate in simulated boiling water reactor (BWR) water on SCC of sensitized Type 304 SS; and (3) radiation-induced segregation (RIS) and irradiation-assisted SCC of Type 304 SS after accumulation of relatively high fluence. Fatigue data were obtained on medium-S-content A533-Gr B and A106-Gr B steels in high-purity (HP) deoxygenated water, in simulated pressurized water reactor (PWR) water, and in air. Crack-growth-rates (CGRs) of composite specimens of A533-Gr B/Inconel-182/Inconel-600 (plated with nickel) and homogeneous specimens of A533-Gr B were determined under small- amplitude cyclic loading in HP water with {approx} 300 ppb dissolved oxygen. CGR tests on sensitized Type 304 SS indicate that low chromate concentrations in BWR water (25--35 ppb) may actually have a beneficial effect on SCC if the sulfate concentration is below a critical level. Microchemical and microstructural changes in HP and commercial-purity Type 304 SS specimens from control-blade absorber tubes used in two operating BWRs were studied by Auger electron spectroscopy and scanning electron microscopy, and slow-strain,rate- tensile tests were conducts on tubular specimens in air and in simulated BWR water at 289{degrees}C.

  7. Ground water quality evaluation near mining area and development of heavy metal pollution index

    NASA Astrophysics Data System (ADS)

    Prasad, Bably; Kumari, Puja; Bano, Shamima; Kumari, Shweta

    2014-03-01

    Opencast as well as underground coal mining are likely to disturb the underground water table in terms of quantity as well as quality. Added to this is the problem of leachates from the large number of industrial waste and overburden dumps that are in abundance in mining areas, reaching the ground water and adversely affecting its quality. Enhancement of heavy metals contamination of the ground water is one eventuality. In the present work, concentrations of 7 heavy metals have been evaluated at 20 important ground water sampling stations at Dhanbad township situated very near to Jharia coalfields. The concentration of heavy metals in general was found to be below the permissible levels although concentration of iron and manganese was found above the permissible limits at a few stations. These data have been used for the calculation of heavy metal pollution index (HPI). The HPI of ground water in total was found to be 6.8860 which is far below the critical index limit of 100 pointing to the fact that the ground water is not polluted with respect to heavy metals in spite of the prolific growth of mining and allied industrial activities near the town.

  8. Ammonium recovery from reject water combined with hydrogen production in a bioelectrochemical reactor.

    PubMed

    Wu, Xue; Modin, Oskar

    2013-10-01

    In this study, a bioelectrochemical reactor was investigated for simultaneous hydrogen production and ammonium recovery from reject water, which is an ammonium-rich side-stream produced from sludge treatment processes at wastewater treatment plants. In the anode chamber of the reactor, microorganisms converted organic material into electrical current. The electrical current was used to generate hydrogen gas at the cathode with 96±6% efficiency. Real or synthetic reject water was fed to the cathode chamber where proton reduction into hydrogen gas resulted in a pH increase which led to ammonium being converted into volatile ammonia. The ammonia could be stripped from the solution and recovered in acid. Overall, ammonium recovery efficiencies reached 94% with synthetic reject water and 79% with real reject water. This process could potentially be used to make wastewater treatment plants more resource-efficient and further research is warranted.

  9. Determination of radioactive elements and heavy metals in sediments and soil from domestic water sources in northern peninsular Malaysia.

    PubMed

    Muhammad, Bashir G; Jaafar, Mohammad Suhaimi; Abdul Rahman, Azhar; Ingawa, Farouk Abdulrasheed

    2012-08-01

    Soil serves as a major reservoir for contaminants as it posseses an ability to bind various chemicals together. To safeguard the members of the public from an unwanted exposure, studies were conducted on the sediments and soil from water bodies that form the major sources of domestic water supply in northern peninsular Malaysia for their trace element concentration levels. Neutron Activation Analysis, using Nigeria Research Reactor-1 (NIRR-1) located at the Centre for Energy Research and Training, Zaria, Nigeria was employed as the analytical tool. The elements identified in major quantities include Na, K, and Fe while As, Br, Cr, U, Th, Eu, Cs, Co, La, Sm, Yb, Sc, Zn, Rb, Ba, Lu, Hf, Ta, and Sb were also identified in trace quantities. Gamma spectroscopy was also employed to analyze some soil samples from the same area. The results indicated safe levels in terms of the radium equivalent activity, external hazard index as well as the mean external exposure dose rates from the soil. The overall screening of the domestic water sources with relatively high heavy metals concentration values in sediments and high activity concentration values in soil is strongly recommended as their accumulation overtime as a consequence of leaching into the water may be of health concern to the members of the public.

  10. Elimination of water pathogens with solar radiation using an automated sequential batch CPC reactor.

    PubMed

    Polo-López, M I; Fernández-Ibáñez, P; Ubomba-Jaswa, E; Navntoft, C; García-Fernández, I; Dunlop, P S M; Schmid, M; Byrne, J A; McGuigan, K G

    2011-11-30

    Solar disinfection (SODIS) of water is a well-known, effective treatment process which is practiced at household level in many developing countries. However, this process is limited by the small volume treated and there is no indication of treatment efficacy for the user. Low cost glass tube reactors, together with compound parabolic collector (CPC) technology, have been shown to significantly increase the efficiency of solar disinfection. However, these reactors still require user input to control each batch SODIS process and there is no feedback that the process is complete. Automatic operation of the batch SODIS process, controlled by UVA-radiation sensors, can provide information on the status of the process, can ensure the required UVA dose to achieve complete disinfection is received and reduces user work-load through automatic sequential batch processing. In this work, an enhanced CPC photo-reactor with a concentration factor of 1.89 was developed. The apparatus was automated to achieve exposure to a pre-determined UVA dose. Treated water was automatically dispensed into a reservoir tank. The reactor was tested using Escherichia coli as a model pathogen in natural well water. A 6-log inactivation of E. coli was achieved following exposure to the minimum uninterrupted lethal UVA dose. The enhanced reactor decreased the exposure time required to achieve the lethal UVA dose, in comparison to a CPC system with a concentration factor of 1.0. Doubling the lethal UVA dose prevented the need for a period of post-exposure dark inactivation and reduced the overall treatment time. Using this reactor, SODIS can be automatically carried out at an affordable cost, with reduced exposure time and minimal user input.

  11. [Pollution Characteristics and Potential Ecological Risk of Heavy Metals in Urban Surface Water Sediments from Yongkang].

    PubMed

    Qi, Peng; Yu, Shu-quan; Zhang, Chao; Liang, Li-cheng; Che, Ji-lu

    2015-12-01

    In order to understand the pollution characteristics of heavy metals in surface water sediments of Yongkang, we analyzed the concentrations of 10 heavy metals including Ti, Cr, Mn, Co, Ni, Cu, Zn, As, Pb and Fe in 122 sediment samples, explored the underlying source of heavy metals and then assessed the potential ecological risks of those metals by methods of the index of geo-accumulation and the potential ecological risk. The study results showed that: 10 heavy metal contents followed the order: Fe > Ti > Mn > Zn > Cr > Cu > Ph > Ni > As > Co, all heavy metals except for Ti were 1. 17 to 3.78 times higher than those of Zhejiang Jinhua- Quzhou basin natural soils background values; The concentrations of all heavy metals had a significantly correlation between each other, indicating that those heavy metals had similar sources of pollution, and it mainly came from industrial and vehicle pollutions; The pollution extent of heavy metals in sediments by geo-accumulation index (Igeo) followed the order: Cr > Zn > Ni > Cu > Fe > As > Pb >Mn > Ti, thereinto, Cr, Zn, Cu and Ni were moderately polluted or heavily polluted at some sampling sites; The potential ecological risk of 9 heavy metals in sediments were in the following order: Cu > As > Ni > Cr > Pb > Co > Zn > Mn > Ti, Cu and As contributed the most to the total potential ecological risk, accounting for 22.84% and 21. 62% , others had a total of 55.54% , through the ecological risk assessment, 89. 34% of the potential ecological risk indexes ( RI) were low and 10. 66% were higher. The contamination level of heavy metals in Yongkang was slight in total, but was heavy in local areas.

  12. Plant control of a fast breeder reactor cooled by supercritical light water

    SciTech Connect

    Nakatsuka, T.; Oka, Y.; Koshizuka, S.

    1997-12-01

    Supercritical water does not exhibit a change of phase. The plant system of the supercritical water cooled reactor is the once-through, direct-cycle where the steam-water separator and coolant recirculation systems are eliminated. It is different from those of BWR and PWR. The reactor is sensitive to the perturbations of the feedwater flow rate, since the whole core coolant driven by the feedwater pumps flows to the turbines. The axial coolant density change is larger than that of a BWR. Pressure control by the feedwater like the supercritical fossil-fired power plant (FPP) is not appropriate because the change of feedwater flow rate largely affects the core power through the coolant density feedback. It is necessary to analyze the controllability of the plant against coolant flow and pressure perturbations for assessing the technical feasibility of the reactor. The plant behaviors of a fast breeder reactor cooled by supercritical water (SCFBR) are analyzed for three principal perturbations: the change of the control rod position, the feedwater flow rate and the turbine control valve opening. Based on the step responses to the perturbations, the plant control system is designed: the pressure is controlled by the turbine control valves, the main steam temperature is controlled by the feedwater flow rate and the core power is controlled by the control rods. Parameters of the control system are selected by the test calculations to satisfy both fast convergence and stability criteria. The plant behaviors with the designed plant control system are stable against the perturbations. The reactor cooled by supercritical light water is controllable with the plant control system designed here. 7 refs., 11 figs., 6 tabs.

  13. Toenail as a biomarker of heavy metal exposure via drinking water: a systematic review.

    PubMed

    Ab Razak, Nurul Hafiza; Praveena, Sarva Mangala; Hashim, Zailina

    2015-01-01

    Toenail is metabolic end product of the skin, which can provide information about heavy metal accumulation in human cells. Slow growth rates of toenail can represent heavy metal exposure from 2 to 12 months before the clipping. The toenail is a non-invasive biomarker that is easy to collect and store and is stable over time. In this systematic review, the suitability of toenail as a long-term biomarker was reviewed, along with the analysis and validation of toenail and confounders to heavy metal. This systematic review has included 30 articles chosen from a total of 132 articles searched from online electronic databases like Pubmed, Proquest, Science Direct, and SCOPUS. Keywords used in the search included "toenail", "biomarker", "heavy metal", and "drinking water". Heavy metal in toenail can be accurately analyzed using an ICP-MS instrument. The validation of toenail heavy metal concentration data is very crucial; however, the Certified Reference Material (CRM) for toenail is still unavailable. Usually, CRM for hair is used in toenail studies. Confounders that have major effects on heavy metal accumulation in toenail are dietary intake of food and supplement, smoking habit, and overall health condition. This review has identified the advantages and limitations of using toenail as a biomarker for long-term exposure, which can help future researchers design a study on heavy metal exposure using toenail.

  14. [Effect of Recycled Water Irrieation on Heavy Metal Pollution in Irrigation Soil].

    PubMed

    Zhou, Yi-qi; Liu, Yun-xia; Fu, Hui-min

    2016-01-15

    With acceleration of urbanization, water shortages will become a serious problem. Usage of reclaimed water for flushing and watering of the green areas will be common in the future. To study the heavy metal contamination of soils after green area irrigation using recycled wastewater from special industries, we selected sewage and laboratory wastewater as water source for integrated oxidation ditch treatment, and the effluent was used as irrigation water of the green area. The irrigation units included broad-leaved forest, bush and lawn. Six samples sites were selected, and 0-20 cm soil of them were collected. Analysis of the heavy metals including Cr, Mn, Ni, Cu, Zn, As, Cd and Pb in the soil showed no significant differences with heavy metals concentration in soil irrigated with tap water. The heavy metals in the soil irrigated with recycled water were mainly enriched in the surface layer, among which the contents of Cr, Ni, Cu, Zn and Pb were below the soil background values of Beijing. A slight pollution of As and Cd was found in the soil irrigated by recycled water, which needs to be noticed.

  15. Reduction of bioavailability and leachability of heavy metals during vermicomposting of water hyacinth.

    PubMed

    Singh, Jiwan; Kalamdhad, Ajay S

    2013-12-01

    Vermicomposting of water hyacinth is a good alternative for the treatment of water hyacinth (Eichhornia crassipes) and subsequentially, beneficial for agriculture purposes. The bioavailability and leachability of heavy metals (Zn, Cu, Mn, Fe, Ni, Pb, Cd, and Cr) were evaluated during vermicomposting of E. crassipes employing Eisenia fetida earthworm. Five different proportions (trials 1, 2, 3, 4, and 5) of cattle manure, water hyacinth, and sawdust were prepared for the vermicomposting process. Results show that very poor biomass growth of earthworms was observed in the highest proportion of water hyacinth (trial 1). The water soluble, diethylenetriaminepentaacetic acid (DTPA) extractable, and leachable heavy metals concentration (percentage of total heavy metals) were reduced significantly in all trials except trial 1. The total concentration of some metals was low but its water soluble and DTPA extractable fractions were similar or more than other metals which were present in higher concentration. This study revealed that the toxicity of metals depends on bioavailable fraction rather than total metal concentration. Bioavailable fraction of metals may be toxic for plants and soil microorganisms. The vermicomposting of water hyacinth by E. fetida was very effective for reduction of bioavailability and leachability of selected heavy metals. Leachability test confirmed that prepared vermicompost is not hazardous for soil, plants, and human health. The feasibility of earthworms to mitigate the metal toxicity and to enhance the nutrient profile in water hyacinth vermicompost might be useful in sustainable land renovation practices at low-input basis.

  16. Trace Analysis of Heavy Metals in Ground Waters of Vijayawada Industrial Area

    ERIC Educational Resources Information Center

    Tadiboyina, Ravisankar; Ptsrk, Prasada Rao

    2016-01-01

    In recent years, the new environmental problem are arising due to industrial hazard wastage, global climate change, ground water contamination and etc., gives an attention to protect environment.one of the major source of contamination of ground water is improper discharge of industrial effluents these effluents contains so many heavy metals which…

  17. Irradiation performance of (Th,Pu)O2 fuel under Pressurized Water Reactor conditions

    NASA Astrophysics Data System (ADS)

    Boer, B.; Lemehov, S.; Wéber, M.; Parthoens, Y.; Gysemans, M.; McGinley, J.; Somers, J.; Verwerft, M.

    2016-04-01

    This paper examines the in-pile safety performance of (Th,Pu)O2 fuel pins under simulated Pressurized Water Reactor (PWR) conditions. Both sol-gel and SOLMAS produced (Th,Pu)O2 fuels at enrichments of 7.9% and 12.8% in Pu/HM have been irradiated at SCK·CEN. The irradiation has been performed under PWR conditions (155 bar, 300 °C) in a dedicated loop of the BR-2 reactor. The loop is instrumented with flow and temperature monitors at inlet and outlet, which allow for an accurate measurement of the deposited enthalpy.

  18. 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.

  19. In-Reactor Oxidation of Zircaloy-4 Under Low Water Vapor Pressures

    SciTech Connect

    Luscher, Walter G.; Senor, David J.; Clayton, Kevin; Longhurst, Glen

    2015-01-01

    Complementary in- and ex-reactor oxidation tests have been performed to evaluate the oxidation and hydrogen absorption performance of Zircaloy-4 (Zr-4) under relatively low partial pressures (300 and 1000 Pa) of water vapor at specified test temperatures (330° and 370°C). Data from these tests will be used to support fabrication of components intended for isotope-producing targets and provide information regarding the temperature and pressure dependence of oxidation and hydrogen absorption of Zr-4 over the specified range of test conditions. Comparisons between in- and ex- reactor test results were performed to evaluate the influence of irradiation.

  20. In-reactor oxidation of zircaloy-4 under low water vapor pressures

    SciTech Connect

    Luscher, Walter G.; Senor, David J.; Clayton, Kevin K.; Longhurst, Glen R.

    2015-01-01

    Complementary in- and ex-reactor oxidation tests have been performed to evaluate the oxidation and hydrogen absorption performance of Zircaloy-4 (Zr-4) under relatively low partial pressures (300 and 1000 Pa) of water vapor at specified test temperatures (330 and 370 ºC). Data from these tests will be used to support the fabrication of components intended for isotope-producing targets and provide information regarding the temperature and pressure dependence of oxidation and hydrogen absorption of Zr- 4 over the specified range of test conditions. Comparisons between in- and ex-reactor test results were performed to evaluate the influence of irradiation.