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Sample records for boiling-water reactors etude

  1. 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)

  2. 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.

  3. 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)

  4. (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.

  5. 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.

  6. 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)

  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. (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.

  9. 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.

  10. 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...

  11. 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)...

  12. 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.

  13. 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.

  14. 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.

  15. 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.

  16. 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)

  17. 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.

  18. 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.

  19. 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)

  20. 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...

  1. 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.

  2. 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)

  3. 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.

  4. 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.

  5. 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.

  6. 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)

  7. 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.

  8. 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.

  9. 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.

  10. 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.

  11. 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.

  12. 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)

  13. 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.

  14. 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

  15. 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.

  16. 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.

  17. 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

  18. 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...

  19. 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...

  20. 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)

  1. 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.

  2. 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)

  3. 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.

  4. 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...

  5. A Compilation of Boiling Water Reactor Operational Experience for the United Kingdom's Office for Nuclear Regulation's Advanced Boiling Water Reactor Generic Design Assessment

    SciTech Connect

    Wheeler, Timothy A.; Liao, Huafei

    2014-12-01

    United States nuclear power plant Licensee Event Reports (LERs), submitted to the United States Nuclear Regulatory Commission (NRC) under law as required by 10 CFR 50.72 and 50.73 were evaluated for reliance to the United Kingdom’s Health and Safety Executive – Office for Nuclear Regulation’s (ONR) general design assessment of the Advanced Boiling Water Reactor (ABWR) design. An NRC compendium of LERs, compiled by Idaho National Laboratory over the time period January 1, 2000 through March 31, 2014, were sorted by BWR safety system and sorted into two categories: those events leading to a SCRAM, and those events which constituted a safety system failure. The LERs were then evaluated as to the relevance of the operational experience to the ABWR design.

  6. 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.

  7. Piping benchmark problems for the General Electric Advanced Boiling Water Reactor

    SciTech Connect

    Bezler, P.; DeGrassi, G.; Braverman, J.; Wang, Y.K.

    1993-08-01

    To satisfy the need for verification of the computer programs and modeling techniques that will be used to perform the final piping analyses for an advanced boiling water reactor standard design, three benchmark problems were developed. The problems are representative piping systems subjected to representative dynamic loads with solutions developed using the methods being proposed for analysis for the advanced reactor standard design. It will be required that the combined license holders demonstrate that their solutions to these problems are in agreement with the benchmark problem set.

  8. 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.

  9. Analysis of the magnetic corrosion product deposits on a boiling water reactor cladding

    SciTech Connect

    Orlov, Andrey; Degueldre, Claude; Kaufmann, Wilfried

    2013-01-15

    The buildup of corrosion product deposits (CRUD) on the fuel cladding of the boiling water reactor (BWR) before and after zinc injection has been investigated by applying local experimental analytical techniques. Under the BWR water chemistry conditions, Zn addition together with the presence of Ni and Mn induce the formation of (Zn,Ni,Mn)[Fe{sub 2}O{sub 4}] spinel solid solutions. X-ray absorption spectroscopy (XAS) revealed inversion ratios of cation distribution in spinels deposited from the solid solution. Based on this information, a two-site ferrite spinel solid solution model is proposed. Electron probe microanalysis (EPMA) and extended X-ray absorption fine structure (EXAFS) findings suggest the zinc-rich ferrite spinels formation on BWR fuel cladding mainly at lower pin. - Graphical Abstract: Analysis of spinels in corrosion product deposits on boiling water reactor fuel rod. Combining EPMA and XAFS results: schematic representation of the ferrite spinels in terms of the end members and their extent of inversion. Note that the ferrites are represented as a surface between the normal (upper plane, M[Fe{sub 2}]O{sub 4}) and the inverse (lower plane, Fe[MFe]O{sub 4}). Actual compositions red Black-Small-Square for the specimen at low elevation (810 mm), blue Black-Small-Square for the specimen at mid elevation (1800 mm). The results have an impact on the properties of the CRUD material. Highlights: Black-Right-Pointing-Pointer Buildup of corrosion product deposits on fuel claddings of a boiling water reactor (BWR) are investigated. Black-Right-Pointing-Pointer Under BWR water conditions, Zn addition with Ni and Mn induced formation of (Zn,Ni,Mn)[Fe{sub 2}O{sub 4}]. Black-Right-Pointing-Pointer X-Ray Adsorption Spectroscopy (XAS) revealed inversion of cations in spinel solid solutions. Black-Right-Pointing-Pointer Zinc-rich ferrite spinels are formed on BWR fuel cladding mainly at lower pin elevations.

  10. 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. Results 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 nondestructive assays.

  11. Instrumentation availability during severe accidents for a boiling water reactor with a Mark I containment

    SciTech Connect

    Arcieri, W.C.; Hanson, D.J. )

    1992-02-01

    In support of the US Nuclear Regulatory Commission Accident Management Research Program, the availability of instruments to supply accident management information during a broad range of severe accidents is evaluated for a Boiling Water Reactor with a Mark I containment. Results from this evaluation include: (1) the identification of plant conditions that would impact instrument performance and information needs during severe accidents; (2) the definition of envelopes of parameters that would be important in assessing the performance of plant instrumentation for a broad range of severe accident sequences; and (3) assessment of the availability of plant instrumentation during severe accidents.

  12. On Stability of Natural-circulation-cooled Boiling Water Reactors during Start-up (Experimental Results)

    SciTech Connect

    Manera, A.; Van der Hagen, T.H.J.J.

    2002-07-01

    The characteristics of flashing-induced instabilities, which are of importance during the start-up phase of natural-circulation Boiling Water Reactors (BWRs), are studied. Experiments at typical start-up conditions (low power and low pressure) are carried out on a steam/water natural circulation loop. The mechanism of flashing-induced instability is analyzed in detail and it is found that non-equilibrium between phases and enthalpy transport plays an important role in the instability process. Pressure and steam volume in the steam dome are found to have a stabilizing effect. The main characteristics of the instabilities have been analyzed. (authors)

  13. Evaluation of the Safety Systems in the Next Generation Boiling Water Reactor

    NASA Astrophysics Data System (ADS)

    Cheng, Ling

    The thesis evaluates the safety systems in the next generation boiling water reactor by analyzing the main steam line break loss of coolant accident performed in the Purdue university multi-dimensional test assembly (PUMA). RELAP5 code simulations, both for the PUMA main steam line break (MSLB) case and for the simplified boiling water reactor (SBWR) MSLB case have been utilized to compare with the experiment data. The comparison shows that RELAP5 is capable to perform the safety analysis for SBWR. The comparison also validates the three-level scaling methodology applied to the design of the PUMA facility. The PUMA suppression pool mixing and condensation test data have been studied to give the detailed understanding on this important local phenomenon. A simple one dimensional integral model, which can reasonably simulate the mixing process inside suppression pool have been developed and the comparison between the model prediction and the experiment data demonstrates the model can be utilized for analyzing the suppression pool mixing process.

  14. Optimal control of a boiling water reactor in load-following via multilevel methods

    SciTech Connect

    Lin, C.; Grossman, L.M.

    1986-04-01

    A multilevel method is applied to the load-following control of a boiling water reactor using a nodal reactor model with practical operational constraints and thermal limits. Due to the very large size of the problem, a decomposition is made using hierarchical control techniques. The optimization of the resulting subproblems is performed using the feasible direction method. An objective functional, of quadratic form, is defined to reflect the control objective, namely to achieve the desired thermal power (tracking) with minimum effort, returning to the initial xenon and iodine concentration as closely as possible. Nodal source equation and discretized Xe-I dynamic equations are formulated as equality constraints, while the linear heat generation rate and the rate of power increase are formulated as inequality constraints. Core flow and control rod position are the control variables. A simplified model of the core is used, with 4 x 4 fuel assemblies that have one control rod at the center.

  15. Passive containment cooling system with drywell pressure regulation for boiling water reactor

    DOEpatents

    Hill, P.R.

    1994-12-27

    A boiling water reactor is described having a regulating valve for placing the wetwell in flow communication with an intake duct of the passive containment cooling system. This subsystem can be adjusted to maintain the drywell pressure at (or slightly below or above) wetwell pressure after the initial reactor blowdown transient is over. This addition to the PCCS design has the benefit of eliminating or minimizing steam leakage from the drywell to the wetwell in the longer-term post-LOCA time period and also minimizes the temperature difference between drywell and wetwell. This in turn reduces the rate of long-term pressure buildup of the containment, thereby extending the time to reach the design pressure limit. 4 figures.

  16. Passive containment cooling system with drywell pressure regulation for boiling water reactor

    DOEpatents

    Hill, Paul R.

    1994-01-01

    A boiling water reactor having a regulating valve for placing the wetwell in flow communication with an intake duct of the passive containment cooling system. This subsystem can be adjusted to maintain the drywell pressure at (or slightly below or above) wetwell pressure after the initial reactor blowdown transient is over. This addition to the PCCS design has the benefit of eliminating or minimizing steam leakage from the drywell to the wetwell in the longer-term post-LOCA time period and also minimizes the temperature difference between drywell and wetwell. This in turn reduces the rate of long-term pressure buildup of the containment, thereby extending the time to reach the design pressure limit.

  17. 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.

  18. 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.

  19. 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.

  20. Aging assessment of the boiling-water reactor (BWR) standby liquid control system. Phase 1

    SciTech Connect

    Orton, R.D.; Johnson, A.B.; Buckley, G.D.; Larson, L.L.

    1992-10-01

    Pacific Northwest Laboratory conducted a Phase I aging assessment of the standby liquid control (SLC) system used in boiling-water reactors. The study was based on detailed reviews of SLC system component and operating experience information obtained from the Nuclear Plant Reliability Database System, the Nuclear Document System, Licensee Event Reports, and other databases. Sources dealing with sodium pentaborate, borates, boric acid, and the effects of environment and corrosion in the SLC system were reviewed to characterize chemical properties and corrosion characteristics of borated solutions. The leading aging degradation concern to date appears to be setpoint drift in relief valves, which has been discovered during routine surveillance and is thought to be caused by mechanical wear. Degradation was also observed in pump seals and internal valves. In general, however, the results of the Phase I study suggest that age-related degradation of SLC systems has not been serious.

  1. Aging assessment of the boiling-water reactor (BWR) standby liquid control system

    SciTech Connect

    Orton, R.D.; Johnson, A.B.; Buckley, G.D.; Larson, L.L.

    1992-10-01

    Pacific Northwest Laboratory conducted a Phase I aging assessment of the standby liquid control (SLC) system used in boiling-water reactors. The study was based on detailed reviews of SLC system component and operating experience information obtained from the Nuclear Plant Reliability Database System, the Nuclear Document System, Licensee Event Reports, and other databases. Sources dealing with sodium pentaborate, borates, boric acid, and the effects of environment and corrosion in the SLC system were reviewed to characterize chemical properties and corrosion characteristics of borated solutions. The leading aging degradation concern to date appears to be setpoint drift in relief valves, which has been discovered during routine surveillance and is thought to be caused by mechanical wear. Degradation was also observed in pump seals and internal valves. In general, however, the results of the Phase I study suggest that age-related degradation of SLC systems has not been serious.

  2. On the interpretation of ECP data from operating boiling water reactors

    SciTech Connect

    Macdonald, D.D.; Urquidi-Macdonald, M.

    1995-12-31

    A method has been devised for estimating the electrochemical conditions that exist in the recirculation piping of a Boiling Water Reactor (BWR) under hydrogen water chemistry (HWC) conditions from corrosion potential measurements that are made in remote autoclaves. The technique makes use of the mixed potential model (MPM) to estimate corrosion potentials in the autoclaves and compares these estimates with measured values in an optimization on the concentrations of hydrogen peroxide and oxygen in the recirculation system. The algorithm recognizes that hydrogen peroxide decomposes in the sampling lines and that the transit times between the recirculation system and the monitoring points depend on the flow rates and sampling line diameters. An analysis is made of ECP data from three monitoring locations in the Barseback BWR in Sweden, as a function of the concentration of hydrogen in the feedwater for two flow rates (5,500 and 6,300 kg/s for the four recirculation loops).

  3. Replacement of outboard main steam isolation valves in a boiling water reactor plant

    SciTech Connect

    Schlereth, J.R.; Pennington, D.

    1996-12-01

    Most Boiling Water Reactor plants utilize wye pattern globe valves for main steam isolation valves for both inboard and outboard isolation. These valves have required a high degree of maintenance attention in order to pass the plant local leakage rate testing (LLRT) requirements at each outage. Northern States Power made a decision in 1993 to replace the outboard valves at it`s Monticello plant with double disc gate valves. The replacement of the outboard valves was completed during the fall outage in 1994. During the spring outage in April of 1996 the first LLRT testing was performed with excellent results. This presentation will address the decision process, time requirements and planning necessary to accomplish the task as well as the performance results and cost effectiveness of replacing these components.

  4. (Installation of a boiling water reactor core melt progression phenomena program)

    SciTech Connect

    Ott, L.J.

    1990-06-07

    The CORA operational staff at Kernforschungszentrum Karlsruhe (KfK) requested, under the auspices of the Severe Fuel Damage Partners Program, that Oak Ridge National Laboratory (ORNL) developed models, specific to boiling water reactor (BWR) response under severe accident conditions, be applied in support of future BWR experiments to be performed in the CORA facility. Accordingly, the current Statement of Work for the BWR Core Melt Progression Phenomena Program provides for the development of a CORA-specific BWR experimental model to analyze the results of CORA BWR experiments and the planning of future experiments. The traveler installed version 1.0 of the CORA/BWR experiment-specific code on KfK personal computers and assisted the CORA staff in their preliminary pretest analyses for CORA test 18.

  5. A decision support system for maintenance management of a boiling-water reactor power plant

    SciTech Connect

    Shen, J.H.; Ray, A.; Levin, S.

    1996-01-01

    This article reports the concept and development of a prototype expert system to serve as a decision support tool for maintenance of boiling-water reactor (BWR) nuclear power plants. The code of the expert system makes use of the database derived from the two BWR units operated by the Pennsylvania Power and Light Company in Berwick, Pennsylvania. The operations and maintenance information from a large number of plant equipment and sub-systems that must be available for emergency conditions and in the event of an accident is stored in the database of the expert system. The ultimate goal of this decision support tool is to identify the relevant Technical Specifications and management rules for shutting down any one of the plant sub-systems or removing a component from service to support maintenance. 6 refs., 7 figs.

  6. Advanced fuel assembly characterization capabilities based on gamma tomography at the Halden boiling water reactor

    SciTech Connect

    Holcombe, S.; Eitrheim, K.; Svaerd, S. J.; Hallstadius, L.; Willman, C.

    2012-07-01

    Characterization of individual fuel rods using gamma spectroscopy is a standard part of the Post Irradiation Examinations performed on experimental fuel at the Halden Boiling Water Reactor. However, due to handling and radiological safety concerns, these measurements are presently carried out only at the end of life of the fuel, and not earlier than several days or weeks after its removal from the reactor core. In order to enhance the fuel characterization capabilities at the Halden facilities, a gamma tomography measurement system is now being constructed, capable of characterizing fuel assemblies on a rod-by-rod basis in a more timely and efficient manner. Gamma tomography for measuring nuclear fuel is based on gamma spectroscopy measurements and tomographic reconstruction techniques. The technique, previously demonstrated on irradiated commercial fuel assemblies, is capable of determining rod-by-rod information without the need to dismantle the fuel. The new gamma tomography system will be stationed close to the Halden reactor in order to limit the need for fuel transport, and it will significantly reduce the time required to perform fuel characterization measurements. Furthermore, it will allow rod-by-rod fuel characterization to occur between irradiation cycles, thus allowing for measurement of experimental fuel repeatedly during its irradiation lifetime. The development of the gamma tomography measurement system is a joint project between the Inst. for Energy Technology - OECD Halden Reactor Project, Westinghouse (Sweden), and Uppsala Univ.. (authors)

  7. 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.

  8. 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.

  9. Knowledge and abilities catalog for nuclear power plant operators: Boiling water reactors, Revision 1

    SciTech Connect

    1995-08-01

    The Knowledge and Abilities Catalog for Nuclear Power Plant Operators: Boiling-Water Reactors (BWRs) (NUREG-1123, Revision 1) provides the basis for the development of content-valid licensing examinations for reactor operators (ROs) and senior reactor operators (SROs). The examinations developed using the BWR Catalog along with the Operator Licensing Examiner Standards (NUREG-1021) and the Examiner`s Handbook for Developing Operator Licensing Written Examinations (NUREG/BR-0122), will cover the topics listed under Title 10, Code of Federal Regulations, Part 55 (10 CFR 55). The BWR Catalog contains approximately 7,000 knowledge and ability (K/A) statements for ROs and SROs at BWRs. The catalog is organized into six major sections: Organization of the Catalog, Generic Knowledge and Ability Statements, Plant Systems grouped by Safety Functions, Emergency and Abnormal Plant Evolutions, Components, and Theory. Revision 1 to the BWR Catalog represents a modification in form and content of the original catalog. The K/As were linked to their applicable 10 CFR 55 item numbers. SRO level K/As were identified by 10 CFR 55.43 item numbers. The plant-wide generic and system generic K/As were combined in one section with approximately one hundred new K/As. Component Cooling Water and Instrument Air Systems were added to the Systems Section. Finally, High Containment Hydrogen Concentration and Plant Fire On Site evolutions added to the Emergency and Abnormal Plant Evolutions section.

  10. Strain-induced corrosion cracking behaviour of low-alloy steels under boiling water reactor conditions

    NASA Astrophysics Data System (ADS)

    Seifert, H. P.; Ritter, S.

    2008-09-01

    The strain-induced corrosion cracking (SICC) behaviour of different low-alloy reactor pressure vessel (RPV) and piping steels and of a RPV weld filler/weld heat-affected zone (HAZ) material was characterized under simulated boiling water reactor (BWR)/normal water chemistry (NWC) conditions by slow rising load (SRL) and very low-frequency fatigue tests with pre-cracked fracture mechanics specimens. Under highly oxidizing BWR/NWC conditions (ECP ⩾+50 mV SHE, ⩾0.4 ppm dissolved oxygen), the SICC crack growth rates were comparable for all materials (hardness <350 HV5) and increased (once initiated) with increasing loading rates and with increasing temperature with a possible maximum/plateau at 250 °C. A minimum KI value of 25 MPa m 1/2 had to be exceeded to initiate SICC in SRL tests. Above this value, the SICC rates increased with increasing loading rate d KI/d t, but were not dependent on the actual KI values up to 60 MPa m 1/2. A maximum in SICC initiation susceptibility occurred at intermediate temperatures around 200-250 °C and at slow strain rates in all materials. In contrast to crack growth, the SICC initiation susceptibility was affected by environmental and material parameters within certain limits.

  11. Characterization of phases in ‘crud’ from boiling-water reactors by transmission electron microscopy

    NASA Astrophysics Data System (ADS)

    Janney, Dawn E.; Porter, Douglas L.

    2007-05-01

    This paper reports phases identified in samples of crud (activated corrosion products) from two commercial boiling-water reactors using transmission and analytical electron microscopy and selected-area electron diffraction. Franklinite (ZnFe 2O 4) was observed in both samples. Hematite (α-Fe 2O 3), crystalline silica (SiO 2), a fine-grained mixture of iron oxides probably including magnetite (Fe 3O 4), hematite (α-Fe 2O 3), and goethite (α-FeOOH), and an unidentified high-Ba, high-S phase were observed in one of the samples. Willemite (Zn 2SiO 4), amorphous silica, and an unidentified iron-chromium phase were observed in the other. Chloride-bearing phases were found in both samples, and are assumed to represent sample contaminants. Because of the small sample volumes and numbers of particles studied and the possibility of contamination, it is not clear whether the differences between the phases observed in the two crud samples represent actual differences in the assemblages formed in the reactors.

  12. Study of plutonium disposition using existing GE advanced Boiling Water Reactors

    SciTech Connect

    Not Available

    1994-06-01

    The end of the cold war and the resulting dismantlement of nuclear weapons has resulted in the need for the US to dispose of 50 to 100 metric tons of excess of plutonium in a safe and proliferation resistant manner. A number of studies, including the recently released National Academy of Sciences (NAS) study, have recommended conversion of plutonium into spent nuclear fuel with its high radiation barrier as the best means of providing permanent conversion and long-term diversion resistance to this material. The NAS study ``Management and Disposition of Excess Weapons Plutonium identified Light Water Reactor spent fuel as the most readily achievable and proven form for the disposition of excess weapons plutonium. The study also stressed the need for a US disposition program which would enhance the prospects for a timely reciprocal program agreement with Russia. This summary provides the key findings of a GE study where plutonium is converted into Mixed Oxide (MOX) fuel and a typical 1155 MWe GE Boiling Water Reactor (BWR) is utilized to convert the plutonium to spent fuel. A companion study of the Advanced BWR has recently been submitted. The MOX core design work that was conducted for the ABWR enabled GE to apply comparable fuel design concepts and consequently achieve full MOX core loading which optimize plutonium throughput for existing BWRs.

  13. A two-step method for developing a control rod program for boiling water reactors

    SciTech Connect

    Taner, M.S.; Levine, S.H. ); Hsiao, M.Y. )

    1992-01-01

    This paper reports on a two-step method that is established for the generation of a long-term control rod program for boiling water reactors (BWRs). The new method assumes a time-variant target power distribution in core depletion. In the new method, the BWR control rod programming is divided into two steps. In step 1, a sequence of optimal, exposure-dependent Haling power distribution profiles is generated, utilizing the spectral shift concept. In step 2, a set of exposure-dependent control rod patterns is developed by using the Haling profiles generated at step 1 as a target. The new method is implemented in a computer program named OCTOPUS. The optimization procedure of OCTOPUS is based on the method of approximation programming, in which the SIMULATE-E code is used to determine the nucleonics characteristics of the reactor core state. In a test in cycle length over a time-invariant, target Haling power distribution case because of a moderate application of spectral shift. No thermal limits of the core were violated. The gain in cycle length could be increased further by broadening the extent of the spetral shift.

  14. Expert system for maintenance management of a boiling water reactor power plant

    SciTech Connect

    Hong Shen; Liou, L.W.; Levine, S.; Ray, A. ); Detamore, M. )

    1992-01-01

    An expert system code has been developed for the maintenance of two boiling water reactor units in Berwick, Pennsylvania, that are operated by the Pennsylvania Power and Light Company (PP and L). The objective of this expert system code, where the knowledge of experienced operators and engineers is captured and implemented, is to support the decisions regarding which components can be safely and reliably removed from service for maintenance. It can also serve as a query-answering facility for checking the plant system status and for training purposes. The operating and maintenance information of a large number of support systems, which must be available for emergencies and/or in the event of an accident, is stored in the data base of the code. It identifies the relevant technical specifications and management rules for shutting down any one of the systems or removing a component from service to support maintenance. Because of the complexity and time needed to incorporate a large number of systems and their components, the first phase of the expert system develops a prototype code, which includes only the reactor core isolation coolant system, the high-pressure core injection system, the instrument air system, the service water system, and the plant electrical system. The next phase is scheduled to expand the code to include all other systems. This paper summarizes the prototype code and the design concept of the complete expert system code for maintenance management of all plant systems and components.

  15. Development of a MELCOR self-initialization algorithm for boiling water reactors

    SciTech Connect

    Chien, C.S.; Wang, S.J.; Cheng, S.K.

    1996-01-01

    The MELCOR code, developed by Sandia National Laboratories, is suitable for calculating source terms and simulating severe accident phenomena of nuclear power plants. Prior to simulating a severe accident transient with MELCOR, the initial steady-state conditions must be generated in advance. The current MELCOR users` manuals do not provide a self-initialization procedure; this is the reason users have to adjust the initial conditions by themselves through a trial-and-error approach. A MELCOR self-initialization algorithm for boiling water reactor plants has been developed, which eliminates the tedious trial-and-error procedures and improves the simulation accuracy. This algorithm adjusts the important plant variable such as the dome pressure, downcomer level, and core flow rate to the desired conditions automatically. It is implemented through input with control functions provided in MELCOR. The reactor power and feedwater temperature are fed as input data. The initialization work of full-power conditions of the Kuosheng nuclear power station is cited as an example. These initial conditions are generated successfully with the developed algorithm. The generated initial conditions can be stored in a restart file and used for transient analysis. The methodology in this study improves the accuracy and consistency of transient calculations. Meanwhile, the algorithm provides all MELCOR users an easy and correct method for establishing the initial conditions.

  16. Optimization of Boiling Water Reactor Loading Pattern Using Two-Stage Genetic Algorithm

    SciTech Connect

    Kobayashi, Yoko; Aiyoshi, Eitaro

    2002-10-15

    A new two-stage optimization method based on genetic algorithms (GAs) using an if-then heuristic rule was developed to generate optimized boiling water reactor (BWR) loading patterns (LPs). In the first stage, the LP is optimized using an improved GA operator. In the second stage, an exposure-dependent control rod pattern (CRP) is sought using GA with an if-then heuristic rule. The procedure of the improved GA is based on deterministic operators that consist of crossover, mutation, and selection. The handling of the encoding technique and constraint conditions by that GA reflects the peculiar characteristics of the BWR. In addition, strategies such as elitism and self-reproduction are effectively used in order to improve the search speed. The LP evaluations were performed with a three-dimensional diffusion code that coupled neutronic and thermal-hydraulic models. Strong axial heterogeneities and constraints dependent on three dimensions have always necessitated the use of three-dimensional core simulators for BWRs, so that optimization of computational efficiency is required. The proposed algorithm is demonstrated by successfully generating LPs for an actual BWR plant in two phases. One phase is only LP optimization applying the Haling technique. The other phase is an LP optimization that considers the CRP during reactor operation. In test calculations, candidates that shuffled fresh and burned fuel assemblies within a reasonable computation time were obtained.

  17. The Neutronics Design and Analysis of a 200-MW(electric) Simplified Boiling Water Reactor Core

    SciTech Connect

    Tinkler, Daniel R.; Downar, Thomas J.

    2003-06-15

    A 200-MW(electric) simplified boiling water reactor (SBWR) was designed and analyzed under sponsorship of the U.S. Department of Energy Nuclear Energy Research Initiative program. The compact size of a 200-MW(electric) reactor makes it attractive for countries with a less well developed engineering infrastructure, as well as for developed countries seeking to tailor generation capacity more closely to the growth of their electricity demand. The 200-MW(electric) core design reported here is based on the 600-MW(electric) General Electric SBWR core, which was first analyzed in the work performed here in order to qualify the computer codes used in the analysis. Cross sections for the 8 x 8 fuel assembly design were generated with the HELIOS lattice physics code, and core simulation was performed with the U.S. Nuclear Regulatory Commission codes RELAP5/PARCS. In order to predict the critical heat flux, the Hench-Gillis correlation was implemented in the RELAP5 code. An equilibrium cycle was designed for the 200-MW(electric) core, which provided a cycle length of more than 2 yr and satisfied the minimum critical power ratio throughout the core life.

  18. Recriticality in a BWR (boiling water reactor) following a core damage event

    SciTech Connect

    Scott, W.B.; Harrison, D.G.; Libby, R.A.; Tokarz, R.D. ); Wooton, R.D.; Denning, R.S.; Tayloe, R.W. Jr. )

    1990-12-01

    This report describes the results of a study conducted by Pacific Northwest Laboratory to assist the US Nuclear Regulatory Commission in evaluating the potential for recriticality in boiling water reactors (BWRs) during certain low probability severe accidents. Based on a conservative bounding analysis, this report concludes that there is a potential for recriticality in BWRs if core reflood occurs after control blade melting has begun but prior to significant fuel rod melting. However, a recriticality event will most likely not generate a pressure pulse significant enough to fail the vessel. Instead, a quasi-steady power level would result and the containment pressure and temperature would increase until the containment failure pressure is reached, unless actions are taken to terminate the event. Two strategies are identified that would aid in regaining control of the reactor and terminate the recriticality event before containment failure pressures are reached. The first strategy involves initiating boration injection at or before the time of core reflood if the potential for control blade melting exists. The second strategy involves initiating residual heat removal suppression pool cooling to remove the heat load generated by the recriticality event and thus extend the time available for boration. 31 figs., 17 tabs.

  19. Spectral measurements of direct and scattered gamma radiation at a boiling-water reactor site

    NASA Astrophysics Data System (ADS)

    Block, R. C.; Preiss, I. L.; Ryan, R. M.; Vargo, G. J.

    1990-12-01

    Quantitative surveys of direct and scattered gamma radiation emitted from the steam-power conversion systems of a boiling-water reactor and other on-site radiation sources were made using a directionally shielded HPGe gamma spectrometry system. The purpose of this study was to obtain data on the relative contributions and energy distributions of direct and scattered gamma radiation in the site environs. The principal radionuclide of concern in this study is 16N produced by the 16O(n,p) 16N reaction in the reactor coolant. Due to changes in facility operation resulting from the implementation of hydrogen water chemistry (HWC), the amount of 16N transported from the reactor to the main steam system under full power operation is excepted to increase by a factor of 1.2 to 5.0. This increase in the 16N source term in the nuclear steam must be considered in the design of new facilities to be constructed on site as well as the evaluation of existing facilities with repect to ALARA (As Low As Reasonably Achievable) dose limits in unrestricted areas. This study consisted of base-line measurements taken under normal BWR chemistry conditions in October, 1987 and a corresponding set taken under HWC conditions in July, 1988. Ground-level and elevated measurements, corresponding to second-story building height, were obtained. The primary conclusion of this study is that direct radiation from the steam-power conversion system is the predominant source of radiation in the site environs of this reactor and that air scattering (i.e. skyshine) does not appear to be significant.

  20. Design-development and operation of the Experimental Boiling-Water Reactor (EBWR) facility, 1955--1967

    SciTech Connect

    Boing, L.E.; Wimunc, E.A.; Whittington, G.A.

    1990-11-01

    The Experimental Boiling-Water Reactor (EBWR) was designed, built, and operated to provide experience and engineering data that would demonstrate the feasibility of the direct-cycle, boiling-water reactor and be applicable to improved, larger nuclear power stations; and was based on information obtained in the first test boiling-water reactors, the BORAX series. EBWR initially produced 20 MW(t), 5 MW(e); later modified and upgraded, as described and illustrated, it was operated at up to 100 MW(t). The facility fulfilled its primary mission -- demonstrating the practicality of the direct-boiling concept -- and, in fact, was the prototype of some of the first commercial plants and of reactor programs in some other countries. After successful completion of the Water-Cooled Reactor Program, EBWR was utilized in the joint Argonne-Hanford Plutonium Recycle Program to develop data for the utilization of plutonium as a fuel in light- water thermal systems. Final shutdown of the EBWR facility followed the termination of the latter program. 13 refs., 12 figs.

  1. Formation and deposition of platinum nanoparticles under boiling water reactor conditions

    NASA Astrophysics Data System (ADS)

    Grundler, Pascal V.; Veleva, Lyubomira; Ritter, Stefan

    2017-10-01

    Stress corrosion cracking (SCC) is a well-known degradation mechanism for components of boiling water reactors (BWRs). Therefore the mitigation of SCC is important for ensuring the integrity of the reactor system. Noble metal chemical application (NMCA) has been developed by General Electric to mitigate SCC and reduce the negative side-effects of hydrogen water chemistry used initially for SCC mitigation. NMCA is now widely applied as an online process (OLNC) during power operation. However, the understanding of the parameters that control the formation and deposition of the noble metal (Pt) particles in a BWR was still incomplete. To fill this knowledge gap, systematic studies on the formation and deposition behaviour of Pt particles in simulated and real BWR environment were performed in the framework of a research project at PSI. The present paper summarizes the most important findings. Experiments in a sophisticated high-temperature water loop revealed that the flow conditions, water chemistry, the Pt injection rate, and the pre-conditioning of the stainless steel surfaces have an impact on the Pt deposition behaviour. Slower Pt injection rates and stoichiometric excess of H2 over O2 produce smaller particles, which may increase the efficiency of the OLNC technique in mitigating SCC. Surfaces with a well-developed oxide layer retain more Pt particles. Furthermore, the pre- and post-OLNC exposure times play an important role for the Pt deposition on specimens exposed at the KKL power plant. Redistribution of Pt in the plant takes place, but most of the Pt apparently does not redeposit on the steel surfaces in the reactor system. Comparison of lab and plant results also demonstrated that plant OLNC applications can be simulated reasonably well on the lab scale.

  2. A bifurcation analysis of boiling water reactor on large domain of parametric spaces

    NASA Astrophysics Data System (ADS)

    Pandey, Vikas; Singh, Suneet

    2016-09-01

    The boiling water reactors (BWRs) are inherently nonlinear physical system, as any other physical system. The reactivity feedback, which is caused by both moderator density and temperature, allows several effects reflecting the nonlinear behavior of the system. Stability analyses of BWR is done with a simplified, reduced order model, which couples point reactor kinetics with thermal hydraulics of the reactor core. The linear stability analysis of the BWR for steady states shows that at a critical value of bifurcation parameter (i.e. feedback gain), Hopf bifurcation occurs. These stable and unstable domains of parametric spaces cannot be predicted by linear stability analysis because the stability of system does not include only stability of the steady states. The stability of other dynamics of the system such as limit cycles must be included in study of stability. The nonlinear stability analysis (i.e. bifurcation analysis) becomes an indispensable component of stability analysis in this scenario. Hopf bifurcation, which occur with one free parameter, is studied here and it formulates birth of limit cycles. The excitation of these limit cycles makes the system bistable in the case of subcritical bifurcation whereas stable limit cycles continues in an unstable region for supercritical bifurcation. The distinction between subcritical and supercritical Hopf is done by two parameter analysis (i.e. codimension-2 bifurcation). In this scenario, Generalized Hopf bifurcation (GH) takes place, which separates sub and supercritical Hopf bifurcation. The various types of bifurcation such as limit point bifurcation of limit cycle (LPC), period doubling bifurcation of limit cycles (PD) and Neimark-Sacker bifurcation of limit cycles (NS) have been identified with the Floquet multipliers. The LPC manifests itself as the region of bistability whereas chaotic region exist because of cascading of PD. This region of bistability and chaotic solutions are drawn on the various

  3. Numerical Evaluation of Fluid Mixing Phenomena in Boiling Water Reactor Using Advanced Interface Tracking Method

    NASA Astrophysics Data System (ADS)

    Yoshida, Hiroyuki; Takase, Kazuyuki

    Thermal-hydraulic design of the current boiling water reactor (BWR) is performed with the subchannel analysis codes which incorporated the correlations based on empirical results including actual-size tests. Then, for the Innovative Water Reactor for Flexible Fuel Cycle (FLWR) core, an actual size test of an embodiment of its design is required to confirm or modify such correlations. In this situation, development of a method that enables the thermal-hydraulic design of nuclear reactors without these actual size tests is desired, because these tests take a long time and entail great cost. For this reason, we developed an advanced thermal-hydraulic design method for FLWRs using innovative two-phase flow simulation technology. In this study, a detailed Two-Phase Flow simulation code using advanced Interface Tracking method: TPFIT is developed to calculate the detailed information of the two-phase flow. In this paper, firstly, we tried to verify the TPFIT code by comparing it with the existing 2-channel air-water mixing experimental results. Secondary, the TPFIT code was applied to simulation of steam-water two-phase flow in a model of two subchannels of a current BWRs and FLWRs rod bundle. The fluid mixing was observed at a gap between the subchannels. The existing two-phase flow correlation for fluid mixing is evaluated using detailed numerical simulation data. This data indicates that pressure difference between fluid channels is responsible for the fluid mixing, and thus the effects of the time average pressure difference and fluctuations must be incorporated in the two-phase flow correlation for fluid mixing. When inlet quality ratio of subchannels is relatively large, it is understood that evaluation precision of the existing two-phase flow correlations for fluid mixing are relatively low.

  4. Interfacing systems LOCAs (Loss of Coolant Accidents) at boiling water reactors

    SciTech Connect

    Chu, Tsong-Lun; Fitzpatrick, R.; Stoyanov, S.

    1987-01-01

    The work presented in this paper was performed by Brookhaven National Laboratory (BNL) in support of Nuclear Regulatory Commission's (NRC) effort towards the resolution of Generic Issue 105 ''Interfacing System Loss of Coolant Accidents (LOCAs) at Boiling Water Reactors (BWRs).'' For BWRs, intersystem LOCA have typically either not been considered in probabilistic risk analyses, or if considered, were judged to contribute little to the risk estimates because of their perceived low frequency of occurrence. However, recent operating experience indicates that the pressure isolation valves (PIVs) in BWRs may not adequately protect against overpressurization of low pressure systems. The objective of this paper is to present the results of a study which analyzed interfacing system LOCA at several BWRs. The BWRs were selected to best represent a spectrum of BWRs in service using industry operating event experience and plant-specific information/configurations. The results presented here include some possible changes in test requirements/practices as well as an evaluation of their reduction potential in terms of core damage frequency (CDF).

  5. Time domain model sensitivity in boiling water reactor stability analysis using TRAC/BF1

    SciTech Connect

    Borkowski, J.A. ); Robinson, G.E.; Baratta, A.J.; Kattic, M. . Dept. of Nuclear Engineering)

    1993-07-01

    Boiling water nuclear reactors (BWRs) may experience density wave instabilities. These instabilities cause the density, and consequently the mass flow rate, to oscillate in the shrouded fuel bundles. This effect causes the nuclear power generation to oscillate because of the tight coupling of flow to power, especially under gravity-driven circulation. To predict the amplitude of the power oscillation, a time domain transient analysis tool may be employed. The modeling tool must have sufficient hydrodynamic detail to model natural circulation in two-phase flow as well as the coupled nuclear feedback. TRAC/BF1 is a modeling code with such capabilities. A dynamic system model is developed for a typical BWR. Using this tool, it is demonstrated that density waves may be modeled in this fashion and that their resultant hydrodynamic and nuclear behavior correspond well to simple theory. Several cases are analyzed using this model, the goal being to determine the coupling between the channel hydrodynamics and the nuclear power. As predicted by others, the two-phase friction controls the extent of the oscillation. Because of this sensitivity, existing conventional methodologies of implementing two-phase friction into analysis codes of this type can lead to significant deviation in results from one case to another. It is found that higher dimensional nuclear feedback models reduce the extent of the oscillation.

  6. Data Reconciliation in the Steam-Turbine Cycle of a Boiling Water Reactor

    SciTech Connect

    Sunde, Svein; Berg, Oivind; Dahlberg, Lennart; Fridqvist, Nils-Olof

    2003-08-15

    A mathematical model for a boiling water reactor steam-turbine cycle was assembled by means of a configurable, steady-state modeling tool TEMPO. The model was connected to live plant data and intermittently fitted to these by minimization of a weighted least-squares object function. The improvement in precision achieved by this reconciliation was assessed from quantities calculated from the model equations linearized around the minimum and from Monte Carlo simulations. It was found that the inclusion of the flow-passing characteristics of the turbines in the model equations significantly improved the precision as compared to simple mass and energy balances, whereas heat transfer calculations in feedwater heaters did not. Under the assumption of linear model equations, the quality of the fit can also be expressed as a goodness-of-fit Q. Typical values for Q were in the order of 0.9. For a validated model Q may be used as a fault detection indicator, and Q dropped to very low values in known cases of disagreement between the model and the plant state. The sensitivity of Q toward measurement faults is discussed in relation to redundancy. The results of the linearized theory and Monte Carlo simulations differed somewhat, and if a more accurate analysis is required, this is better based on the latter. In practical application of the presently employed techniques, however, assessment of uncertainties in raw data is an important prerequisite.

  7. Many-Group Cross-Section Adjustment Techniques for Boiling Water Reactor Adaptive Simulation

    SciTech Connect

    Jessee, Matthew Anderson

    2011-01-01

    Computational capability has been developed to adjust multigroup neutron cross sections, including self-shielding correction factors, to improve the fidelity of boiling water reactor (BWR) core modeling and simulation. The method involves propagating multigroup neutron cross-section uncertainties through various BWR computational models to evaluate uncertainties in key core attributes such as core k{sub eff}, nodal power distributions, thermal margins, and in-core detector readings. Uncertainty-based inverse theory methods are then employed to adjust multigroup cross sections to minimize the disagreement between BWR core modeling predictions and observed (i.e., measured) plant data. For this paper, observed plant data are virtually simulated in the form of perturbed three-dimensional nodal power distributions with the perturbations sized to represent actual discrepancies between predictions and real plant data. The major focus of this work is to efficiently propagate multigroup neutron cross-section uncertainty through BWR lattice physics and core simulator calculations. The data adjustment equations are developed using a subspace approach that exploits the ill-conditioning of the multigroup cross-section covariance matrix to minimize computation and storage burden. Tikhonov regularization is also employed to improve the conditioning of the data adjustment equations. Expressions are also provided for posterior covariance matrices of both the multigroup cross-section and core attributes uncertainties.

  8. Technology, safety and costs of decommissioning a Reference Boiling Water Reactor Power Station. Main report. Volume 1

    SciTech Connect

    Oak, H.D.; Holter, G.M.; Kennedy, W.E. Jr.; Konzek, G.J.

    1980-06-01

    Technology, safety and cost information is given for the conceptual decommissioning of a large (1100MWe) boiling water reactor (BWR) power station. Three approaches to decommissioning, immediate dismantlement, safe storage with deferred dismantlement and entombment, were studied to obtain comparisons between costs, occupational radiation doses, potential dose to the public and other safety impacts. It also shows the sensitivity of decommissioning safety and costs to the power rating of a BWR in the range of 200 to 1100 MWe.

  9. Technology, safety and costs of decommissioning a reference boiling water reactor power station. Classification of decommissioning wastes. Addendum 2

    SciTech Connect

    Murphy, E.S.

    1984-09-01

    The radioactive wastes expected to result from decommissioning of the reference boiling water reactor power station are reviewed and classified in accordance with 10 CFR 61. The 18,949 cubic meters of waste from DECON are classified as follows: Class A, 97.5%; Class B, 2.0%; Class C, 0.3%. About 0.2% (47 cubic meters) of the waste would be generally unacceptable for disposal using near-surface disposal methods.

  10. Study of plutonium disposition using the GE Advanced Boiling Water Reactor (ABWR)

    SciTech Connect

    1994-04-30

    The end of the cold war and the resulting dismantlement of nuclear weapons has resulted in the need for the U.S. to disposition 50 to 100 metric tons of excess of plutonium in parallel with a similar program in Russia. A number of studies, including the recently released National Academy of Sciences (NAS) study, have recommended conversion of plutonium into spent nuclear fuel with its high radiation barrier as the best means of providing long-term diversion resistance to this material. The NAS study {open_quotes}Management and Disposition of Excess Weapons Plutonium{close_quotes} identified light water reactor spent fuel as the most readily achievable and proven form for the disposition of excess weapons plutonium. The study also stressed the need for a U.S. disposition program which would enhance the prospects for a timely reciprocal program agreement with Russia. This summary provides the key findings of a GE study where plutonium is converted into Mixed Oxide (MOX) fuel and a 1350 MWe GE Advanced Boiling Water Reactor (ABWR) is utilized to convert the plutonium to spent fuel. The ABWR represents the integration of over 30 years of experience gained worldwide in the design, construction and operation of BWRs. It incorporates advanced features to enhance reliability and safety, minimize waste and reduce worker exposure. For example, the core is never uncovered nor is any operator action required for 72 hours after any design basis accident. Phase 1 of this study was documented in a GE report dated May 13, 1993. DOE`s Phase 1 evaluations cited the ABWR as a proven technical approach for the disposition of plutonium. This Phase 2 study addresses specific areas which the DOE authorized as appropriate for more in-depth evaluations. A separate report addresses the findings relative to the use of existing BWRs to achieve the same goal.

  11. 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.

  12. MODULAR AND FULL SIZE SIMPLIFIED BOILING WATER REACTOR DESIGN WITH FULLY PASSIVE SAFETY SYSTEMS

    SciTech Connect

    M. Ishii; S. T. Revankar; T. Downar; Y. Xu, H. J. Yoon; D. Tinkler; U. S. Rohatgi

    2003-06-16

    OAK B204 The overall goal of this three-year research project was to develop a new scientific design of a compact modular 200 MWe and a full size 1200 MWe simplified boiling water reactors (SBWR). Specific objectives of this research were: (1) to perform scientific designs of the core neutronics and core thermal-hydraulics for a small capacity and full size simplified boiling water reactor, (2) to develop a passive safety system design, (3) improve and validate safety analysis code, (4) demonstrate experimentally and analytically all design functions of the safety systems for the design basis accidents (DBA) and (5) to develop the final scientific design of both SBWR systems, 200 MWe (SBWR-200) and 1200 MWe (SBWR-1200). The SBWR combines the advantages of design simplicity and completely passive safety systems. These advantages fit well within the objectives of NERI and the Department of Energy's focus on the development of Generation III and IV nuclear power. The 3-year research program was structured around seven tasks. Task 1 was to perform the preliminary thermal-hydraulic design. Task 2 was to perform the core neutronic design analysis. Task 3 was to perform a detailed scaling study and obtain corresponding PUMA conditions from an integral test. Task 4 was to perform integral tests and code evaluation for the DBA. Task 5 was to perform a safety analysis for the DBA. Task 6 was to perform a BWR stability analysis. Task 7 was to perform a final scientific design of the compact modular SBWR-200 and the full size SBWR-1200. A no cost extension for the third year was requested and the request was granted and all the project tasks were completed by April 2003. The design activities in tasks 1, 2, and 3 were completed as planned. The existing thermal-hydraulic information, core physics, and fuel lattice information was collected on the existing design of the simplified boiling water reactor. The thermal-hydraulic design were developed. Based on a detailed integral

  13. Factors influencing the precoat filtration of boiling water reactor water streams

    SciTech Connect

    Hermansson, H.P. ); Persson, G. ); Reinvall, A. )

    1994-10-01

    A series of studies on precoat filtration were carried out on condensate and preheater drains in the Swedish and Finnish boiling water reactors (BWRs). The goal was to increase knowledge about the precoat filtration process and to find physical and chemical means to improve the performance of the precoat filters in the condensate polishing plants. To achieve this goal a number of parameters, such as type of resin, bed depth, pH, oxygen and organic contaminant concentrations (measured total organic carbon), and corrosion product particle characteristics, were selected for the study. The work was mainly carried out in the power plants using an experimental facility fed with on-line sampled condensates and drains taken from the plant sampling lines. The main results are that there is a varying influence on precoat filtration from all the aforementioned parameters. The oxygen concentration, the concentration of organic contaminants, and the type of corrosion products are, however, the factors that have the strongest influence within the parameter ranges that are representative for BWR operation. The results are rather similar when the different units are compared. There are, however, some differences that could be mainly attributed to deviations in operation parameters and the subsequent differences in the corrosion product spectra. The mechanism for precoat filtration of corrosion products in BWR condensate is complex. The filtration behavior is to a large extent governed by competition between depth filtration and electrostatic interactions. During the early stages of the filtration cycle, electrostatic interaction is of great importance, whereas depth filtration becomes more important with increasing operating time. Rapid pressure drop buildup rates have been demonstrated to be caused by the presence of amorphous corrosion products. An effect from the presence of organic contaminants has been found, although this should be of little significance.

  14. 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.

  15. Branch-and-Bound algorithm applied to uncertainty quantification of a Boiling Water Reactor Station Blackout

    SciTech Connect

    Nielsen, Joseph; Tokuhiro, Akira; Hiromoto, Robert; Tu, Lei

    2015-11-13

    combinatorial explosion using a Branch-and-Bound algorithm applied to Dynamic Event Trees (DET), which utilize LENDIT (L – Length, E – Energy, N – Number, D – Distribution, I – Information, and T – Time) as well as a set theory to describe system, state, resource, and response (S2R2) sets to create bounding functions for the DET. The optimization of the DET in identifying high probability failure branches is extended to create a Phenomenological Identification and Ranking Table (PIRT) methodology to evaluate modeling parameters important to safety of those failure branches that have a high probability of failure. The PIRT can then be used as a tool to identify and evaluate the need for experimental validation of models that have the potential to reduce risk. Finally, in order to demonstrate this methodology, a Boiling Water Reactor (BWR) Station Blackout (SBO) case study is presented.

  16. Branch-and-Bound algorithm applied to uncertainty quantification of a Boiling Water Reactor Station Blackout

    DOE PAGES

    Nielsen, Joseph; Tokuhiro, Akira; Hiromoto, Robert; ...

    2015-11-13

    combinatorial explosion using a Branch-and-Bound algorithm applied to Dynamic Event Trees (DET), which utilize LENDIT (L – Length, E – Energy, N – Number, D – Distribution, I – Information, and T – Time) as well as a set theory to describe system, state, resource, and response (S2R2) sets to create bounding functions for the DET. The optimization of the DET in identifying high probability failure branches is extended to create a Phenomenological Identification and Ranking Table (PIRT) methodology to evaluate modeling parameters important to safety of those failure branches that have a high probability of failure. The PIRT can then be used as a tool to identify and evaluate the need for experimental validation of models that have the potential to reduce risk. Finally, in order to demonstrate this methodology, a Boiling Water Reactor (BWR) Station Blackout (SBO) case study is presented.« less

  17. Comparing Simulation Results with Traditional PRA Model on a Boiling Water Reactor Station Blackout Case Study

    SciTech Connect

    Zhegang Ma; Diego Mandelli; Curtis Smith

    2011-07-01

    A previous study used RELAP and RAVEN to conduct a boiling water reactor station black-out (SBO) case study in a simulation based environment to show the capabilities of the risk-informed safety margin characterization methodology. This report compares the RELAP/RAVEN simulation results with traditional PRA model results. The RELAP/RAVEN simulation run results were reviewed for their input parameters and output results. The input parameters for each simulation run include various timing information such as diesel generator or offsite power recovery time, Safety Relief Valve stuck open time, High Pressure Core Injection or Reactor Core Isolation Cooling fail to run time, extended core cooling operation time, depressurization delay time, and firewater injection time. The output results include the maximum fuel clad temperature, the outcome, and the simulation end time. A traditional SBO PRA model in this report contains four event trees that are linked together with the transferring feature in SAPHIRE software. Unlike the usual Level 1 PRA quantification process in which only core damage sequences are quantified, this report quantifies all SBO sequences, whether they are core damage sequences or success (i.e., non core damage) sequences, in order to provide a full comparison with the simulation results. Three different approaches were used to solve event tree top events and quantify the SBO sequences: “W” process flag, default process flag without proper adjustment, and default process flag with adjustment to account for the success branch probabilities. Without post-processing, the first two approaches yield incorrect results with a total conditional probability greater than 1.0. The last approach accounts for the success branch probabilities and provides correct conditional sequence probabilities that are to be used for comparison. To better compare the results from the PRA model and the simulation runs, a simplified SBO event tree was developed with only four

  18. 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.

  19. Design of a boiling water reactor equilibrium core using thorium-uranium fuel

    SciTech Connect

    Francois, J-L.; Nunez-Carrera, A.; Espinosa-Paredes, G.; Martin-del-Campo, C.

    2004-10-06

    In this paper the design of a Boiling Water Reactor (BWR) equilibrium core using thorium is presented; a heterogeneous blanket-seed core arrangement concept was adopted. The design was developed in three steps: in the first step two different assemblies were designed based on the integrated blanket-seed concept, they are the blanket-dummy assembly and the blanket-seed assembly. The integrated blanketseed concept comes from the fact that the blanket and the seed rods are located in the same assembly, and are burned-out in a once-through cycle. In the second step, a core design was developed to achieve an equilibrium cycle of 365 effective full power days in a standard BWR with a reload of 104 fuel assemblies designed with an average 235U enrichment of 7.5 w/o in the seed sub-lattice. The main operating parameters, like power, linear heat generation rate and void distributions were obtained as well as the shutdown margin. It was observed that the analyzed parameters behave like those obtained in a standard BWR. The shutdown margin design criterion was fulfilled by addition of a burnable poison region in the assembly. In the third step an in-house code was developed to evaluate the thorium equilibrium core under transient conditions. A stability analysis was also performed. Regarding the stability analysis, five operational states were analyzed; four of them define the traditional instability region corner of the power-flow map and the fifth one is the operational state for the full power condition. The frequency and the boiling length were calculated for each operational state. The frequency of the analyzed operational states was similar to that reported for BWRs; these are close to the unstable region that occurs due to the density wave oscillation phenomena in some nuclear power plants. Four transient analyses were also performed: manual SCRAM, recirculation pumps trip, main steam isolation valves closure and loss of feed water. The results of these transients are

  20. Acoustic Analysis for a Steam Dome and Pipings of a 1,100 MWe-Class Boiling Water Reactor

    SciTech Connect

    Yasumi, Kitajima; Masanobu, Watanabe; Keiji, Matsunaga; Tsuyoshi, Hagiwara

    2006-07-01

    For the integrity evaluation of steam dryers in up-rated nuclear power plants, we have applied acoustic analysis to a nuclear power plant steam dome and main steam pipings. We have selected a 1,100 MWe-class boiling water reactor as a subject of the analysis. We have constructed a three-dimensional finite element model, and conducted acoustic analyses. The analysis result suggested that the origin of steam pressure pulsation in high frequency range was due to vortex shedding at standpipes. (authors)

  1. Preliminary phenomena identification and ranking tables for simplified boiling water reactor Loss-of-Coolant Accident scenarios

    SciTech Connect

    Kroeger, P.G.; Rohatgi, U.S.; Jo, J.H.; Slovik, G.C.

    1998-04-01

    For three potential Loss-of-Coolant Accident (LOCA) scenarios in the General Electric Simplified Boiling Water Reactors (SBWR) a set of Phenomena Identification and Ranking Tables (PIRT) is presented. The selected LOCA scenarios are typical for the class of small and large breaks generally considered in Safety Analysis Reports. The method used to develop the PIRTs is described. Following is a discussion of the transient scenarios, the PIRTs are presented and discussed in detailed and in summarized form. A procedure for future validation of the PIRTs, to enhance their value, is outlined. 26 refs., 25 figs., 44 tabs.

  2. Design and Analysis of Thorium-fueled Reduced Moderation Boiling Water Reactors

    NASA Astrophysics Data System (ADS)

    Gorman, Phillip Michael

    The Resource-renewable Boiling Water Reactors (RBWRs) are a set of light water reactors (LWRs) proposed by Hitachi which use a triangular lattice and high void fraction to incinerate fuel with an epithermal spectrum, which is highly atypical of LWRs. The RBWRs operate on a closed fuel cycle, which is impossible with a typical thermal spectrum reactor, in order to accomplish missions normally reserved for sodium fast reactors (SFRs)--either fuel self-sufficiency or waste incineration. The RBWRs also axially segregate the fuel into alternating fissile "seed" regions and fertile "blanket" regions in order to enhance breeding and leakage probability upon coolant voiding. This dissertation focuses on thorium design variants of the RBWR: the self-sufficient RBWR-SS and the RBWR-TR, which consumes reprocessed transuranic (TRU) waste from PWR used nuclear fuel. These designs were based off of the Hitachi-designed RBWR-AC and the RBWR-TB2, respectively, which use depleted uranium (DU) as the primary fertile fuel. The DU-fueled RBWRs use a pair of axially segregated seed sections in order to achieve a negative void coefficient; however, several concerns were raised with this multi-seed approach, including difficulty with controlling the reactor and unacceptably high axial power peaking. Since thorium-uranium fuel tends to have much more negative void feedback than uranium-plutonium fuels, the thorium RBWRs were designed to use a single elongated seed to avoid these issues. A series of parametric studies were performed in order to find the design space for the thorium RBWRs, and optimize the designs while meeting the required safety constraints. The RBWR-SS was optimized to maximize the discharge burnup, while the RBWR-TR was optimized to maximize the TRU transmutation rate. These parametric studies were performed on an assembly level model using the MocDown simulator, which calculates an equilibrium fuel composition with a specified reprocessing scheme. A full core model was

  3. Modeling and numerical simulation of oscillatory two-phase flows, with application to boiling water nuclear reactors

    SciTech Connect

    Rosa, M.P.; Podowski, M.Z.

    1995-09-01

    This paper is concerned with the analysis of dynamics and stability of boiling channels and systems. The specific objectives are two-fold. One of them is to present the results of a study aimed at analyzing the effects of various modeling concepts and numerical approaches on the transient response and stability of parallel boiling channels. The other objective is to investigate the effect of closed-loop feedback on stability of a boiling water reactor (BWR). Various modeling and computational issues for parallel boiling channels are discussed, such as: the impact of the numerical discretization scheme for the node containing the moving boiling boundary on the convergence and accuracy of computations, and the effects of subcooled boiling and other two-phase flow phenomena on the predictions of marginal stability conditions. Furthermore, the effects are analyzed of local loss coefficients around the recirculation loop of a boiling water reactor on stability of the reactor system. An apparent paradox is explained concerning the impact of changing single-phase losses on loop stability. The calculations have been performed using the DYNOBOSS computer code. The results of DYNOBOSS validation against other computer codes and experimental data are shown.

  4. Large-Break Loss-of-Coolant Accident Testing and Simulation for 200-MWe Simplified Boiling Water Reactor

    SciTech Connect

    Revankar, S.T.; Xu, Y.; Yoon, H.J.; Ishii, M.

    2002-07-01

    The performance of the safety systems of a new design of the 200-MWe simplified boiling water reactor during a large-break, loss-of-coolant accident transient was investigated through code modeling and integral system testing. The accident considered was a break in the main steam line which is the major design basis accident. RELAP5/MOD3 best estimate reactor thermalhydraulic code was used and its applicability to the reactor safety system evaluation was examined. The integral tests were performed to assess the safety systems and the response of the emergency core cooling systems to accident conditions in a scaled facility called PUMA. The details of the safety system behavior are presented. The integral test simulations examined code applicability at the scaled facility level as well as prototype key safety system performance. (authors)

  5. Final safety evaluation report related to the certification of the advanced boiling water reactor design. Volume 1: Main report

    SciTech Connect

    Not Available

    1994-07-01

    This safety evaluation report (SER) documents the technical review of the US Advanced Boiling Water Reactor (ABWR) standard design by the US Nuclear Regulatory Commission (NRC) staff. The application for the ABWR design was initially submitted by the General Electric Company, now GE Nuclear Energy (GE), in accordance with the procedures of Appendix O of Part 50 of Title 10 of the Code of Federal Regulations (10 CFR Part 50). Later GE requested that its application be considered as an application for design approval and subsequent design certification pursuant to 10 CFR {section} 52.45. The ABWR is a single-cycle, forced-circulation, boiling water reactor (BWR) with a rated power of 3,926 megawatts thermal (MWt) and a design power of 4,005 MWt. To the extent feasible and appropriate, the staff relied on earlier reviews for those ABWR design features that are substantially the same as those previously considered. Unique features of the ABWR design include internal recirculation pumps, fine-motion control rod drives, microprocessor-based digital logic and control systems, and digital safety systems. On the basis of its evaluation and independent analyses, the NRC staff concludes that, subject to satisfactory resolution of the confirmatory items identified in Section 1.8 of this SER, GE`s application for design certification meets the requirements of Subpart B of 10 CFR Part 52 that are applicable and technically relevant to the US ABWR standard design.

  6. Final safety evaluation report related to the certification of the advanced boiling water reactor design. Volume 2: Appendices

    SciTech Connect

    Not Available

    1994-07-01

    This safety evaluation report (SER) documents the technical review of the US Advanced Boiling Water Reactor (ABWR) standard design by the US Nuclear Regulatory Commission (NRC) staff. The application for the ABWR design was initially submitted by the General Electric Company, now GE Nuclear Energy (GE), in accordance with the procedures of Appendix O of Part 50 of Title 10 of the code of Federal Regulations (10 CFR Part 50). Later GE requested that its application be considered as an application for design approval and subsequent design certification pursuant to 10 CFR {section} 52.45. The ABWR is a single-cycle, forced-circulation, boiling water reactor (BWR) with a rated power of 3,926 megawatts thermal (MWt) and a design power of 4,005 MWt. To the extent feasible and appropriate, the staff relied on earlier reviews for those ABWR design features that are substantially the same as those previously considered. Unique features of the ABWR design include internal recirculation pumps, fine-motion control rod drives, microprocessor-based digital logic and control systems, and digital safety systems. On the basis of its evaluation and independent analyses, the NRC staff concludes that, subject to satisfactory resolution of the confirmatory items identified in Section 1.8 of this SER, GE`s application for design certification meets the requirements of Subpart B of 10 CFR Part 52 that are applicable and technically relevant to the US ABWR standard design.

  7. Experimental and Thermalhydraulic Code Assessment of the Transient Behavior of the Passive Condenser System in an Advanced Boiling Water Reactor

    SciTech Connect

    S.T. Revankar; W. Zhou; Gavin Henderson

    2008-07-08

    The main goal of the project was to study analytically and experimentally the condensation heat transfer for the passive condenser system such as GE Economic Simplified Boiling Water Reactor (ESBWR). The effect of noncondensable gas in condenser tube and the reduction of secondary pool water level to the condensation heat transfer coefficient was the main focus in this research. The objectives of this research were to : 1) obtain experimental data on the local and tube averaged condensation heat transfer rates for the PCCS with non-condensable and with change in the secondary pool water, 2) assess the RELAP5 and TRACE computer code against the experimental data, and 3) develop mathematical model and ehat transfer correlation for the condensation phenomena for system code application. The project involves experimentation, theoretical model development and verification, and thermal- hydraulic codes assessment.

  8. Analytic solution to verify code predictions of two-phase flow in a boiling water reactor core channel

    SciTech Connect

    Chen, K.F.; Olson, C.A.

    1983-09-01

    One reliable method that can be used to verify the solution scheme of a computer code is to compare the code prediction to a simplified problem for which an analytic solution can be derived. An analytic solution for the axial pressure drop as a function of the flow was obtained for the simplified problem of homogeneous equilibrium two-phase flow in a vertical, heated channel with a cosine axial heat flux shape. This analytic solution was then used to verify the predictions of the CONDOR computer code, which is used to evaluate the thermal-hydraulic performance of boiling water reactors. The results show excellent agreement between the analytic solution and CONDOR prediction.

  9. Analytic solution to verify code predictions of two-phase flow in a boiling water reactor core channel. [CONDOR code

    SciTech Connect

    Chen, K.F.; Olson, C.A.

    1983-09-01

    One reliable method that can be used to verify the solution scheme of a computer code is to compare the code prediction to a simplified problem for which an analytic solution can be derived. An analytic solution for the axial pressure drop as a function of the flow was obtained for the simplified problem of homogeneous equilibrium two-phase flow in a vertical, heated channel with a cosine axial heat flux shape. This analytic solution was then used to verify the predictions of the CONDOR computer code, which is used to evaluate the thermal-hydraulic performance of boiling water reactors. The results show excellent agreement between the analytic solution and CONDOR prediction.

  10. Preliminary design study of small long life boiling water reactor (BWR) with tight lattice thorium nitride fuel

    NASA Astrophysics Data System (ADS)

    Trianti, Nuri; Su'ud, Zaki; Arif, Idam; Riyana, EkaSapta

    2014-09-01

    Neutronic performance of small long-life boiling water reactors (BWR) with thorium nitride based fuel has been performed. A recent study conducted on BWR in tight lattice environments (with a lower moderator percentage) produces small power reactor which has some specifications, i.e. 10 years operation time, power density of 19.1 watt/cc and maximum excess reactivity of about 4%. This excess reactivity value is smaller than standard reactivity of conventional BWR. The use of hexagonal geometry on the fuel cell of BWR provides a substantial effect on the criticality of the reactor to obtain a longer operating time. Supported by a tight concept lattice where the volume fraction of the fuel is greater than the moderator and fuel, Thorium Nitride give good results for fuel cell design on small long life BWR. The excess reactivity of the reactor can be reduced with the addition of gadolinium as burnable poisons. Therefore the hexagonal tight lattice fuel cell design of small long life BWR that has a criticality more than 20 years of operating time has been obtained.

  11. Preliminary design study of small long life boiling water reactor (BWR) with tight lattice thorium nitride fuel

    SciTech Connect

    Trianti, Nuri E-mail: szaki@fi.itba.c.id; Su'ud, Zaki E-mail: szaki@fi.itba.c.id; Arif, Idam E-mail: szaki@fi.itba.c.id; Riyana, EkaSapta

    2014-09-30

    Neutronic performance of small long-life boiling water reactors (BWR) with thorium nitride based fuel has been performed. A recent study conducted on BWR in tight lattice environments (with a lower moderator percentage) produces small power reactor which has some specifications, i.e. 10 years operation time, power density of 19.1 watt/cc and maximum excess reactivity of about 4%. This excess reactivity value is smaller than standard reactivity of conventional BWR. The use of hexagonal geometry on the fuel cell of BWR provides a substantial effect on the criticality of the reactor to obtain a longer operating time. Supported by a tight concept lattice where the volume fraction of the fuel is greater than the moderator and fuel, Thorium Nitride give good results for fuel cell design on small long life BWR. The excess reactivity of the reactor can be reduced with the addition of gadolinium as burnable poisons. Therefore the hexagonal tight lattice fuel cell design of small long life BWR that has a criticality more than 20 years of operating time has been obtained.

  12. Design study of an irradiation experiment with inert matrix and mixed-oxide fuel at the Halden boiling water reactor

    NASA Astrophysics Data System (ADS)

    Kasemeyer, U.; Joo, H.-K.; Ledergerber, G.

    1999-08-01

    An effective way to reduce the large quantities of plutonium currently accumulated worldwide would be to use uranium-free fuel in light water reactors (LWRs) so that no new plutonium is produced. To test such a new fuel under reactor conditions and in comparison with standard mixed-oxide (MOX) fuel, an irradiation experiment is planned at the Halden boiling water reactor. The behaviour of three fuel rods consisting of uranium-free fuel will be investigated together with three rods made out of uranium-plutonium mixed-oxide fuel in the same assembly. The fuel compositions were adjusted so that all rods produce a similar power. Because of the moderation with D 2O in the Halden reactor, two different surroundings of the considered assembly were examined to analyze the influence of the flux spectrum on the experiment. This showed that the influence of the spectrum on the material behaviour is negligible. The relation between assembly power and average neutron detector signal as well as the burnup or depletion function was calculated. The assumed power history was adapted to a usual LWR schedule. It is possible to reach a burnup of ˜540 MW d kg HM-1 with the uranium-free fuel and ˜54 MW d kg HM-1 with the MOX fuel after five years of irradiation, which is similar to the average burnup reached in commercial LWRs after four years of operation.

  13. Coupled thermohydraulic-neutronic instabilities in boiling water nuclear reactors: A review of the state of the art

    SciTech Connect

    March-Leuba, J. ); Rey, J.M. )

    1992-01-01

    This paper provides a review of the current state of the art on the topic of coupled neutronic-thermohydraulic instabilities in boiling water nuclear reactors (BWRs). The topic of BWR instabilities is of great current relevance since it affects the operation of a large number of commercial nuclear reactors. The recent trends towards introduction of high efficiency fuels that permit reactor operation at higher power densities with increased void reactivity feedback and decreased response times, has resulted in a decrease of the stability margin in the low-flow, high-power region of the operating map. This trend has resulted in a number of unexpected'' instability events. For instance, United States plants have experienced two instability events recently, one of them resulted in an automatic reactor scram; in Spain, two BWR plants have experienced unstable limit cycle oscillations that required operator action to suppress. Similar events have been experienced in other European countries. In recent years, BWR instabilities have been one of the more exciting topics of work in the area of transient thermohydraulics. As a result, significant advances in understanding the physics behind these events have occurred, and a new and improved'' state of the art has emerged recently.

  14. Coupled thermohydraulic-neutronic instabilities in boiling water nuclear reactors: A review of the state of the art

    SciTech Connect

    March-Leuba, J.; Rey, J.M.

    1992-05-01

    This paper provides a review of the current state of the art on the topic of coupled neutronic-thermohydraulic instabilities in boiling water nuclear reactors (BWRs). The topic of BWR instabilities is of great current relevance since it affects the operation of a large number of commercial nuclear reactors. The recent trends towards introduction of high efficiency fuels that permit reactor operation at higher power densities with increased void reactivity feedback and decreased response times, has resulted in a decrease of the stability margin in the low-flow, high-power region of the operating map. This trend has resulted in a number of ``unexpected`` instability events. For instance, United States plants have experienced two instability events recently, one of them resulted in an automatic reactor scram; in Spain, two BWR plants have experienced unstable limit cycle oscillations that required operator action to suppress. Similar events have been experienced in other European countries. In recent years, BWR instabilities have been one of the more exciting topics of work in the area of transient thermohydraulics. As a result, significant advances in understanding the physics behind these events have occurred, and a ``new and improved`` state of the art has emerged recently.

  15. A proof-of-concept transient diagnostic expert system for BWRs (Boiling Water Reactors)

    SciTech Connect

    Yoshida, K.; Naser, J.A.

    1988-05-01

    A proof-of-concept transient diagnostic expert system has been developed to identify the cause and the type of an abnormal transient in a boiling water nuclear power plant. For this expert system development, the calculational results of the simulation code RETRAN were used as the knowledge source. The knowledge extracted from the RETRAN analyses was transformed into IF-THEN rules in the knowledge base for the expert system. An important feature of this expert system is the introduction of certainty factors to allow diagnosis even in the cases where data may be either missing or marked as invalid. To increase the capability of this diagnostic system to distinguish between similiar transients, backward chaining reasoning is used to support the forward chaining reasoning with certainty factors. Through this effort, it has been demonstrated that an expert system can be successfully used to create a transient diagnostic system. It has also successfully demonstrated that RETRAN can be used as the knowledge source for developing the knowledge base of the diagnostic system.

  16. Analysis of cracked core spray injection line piping from the Quad Cities Units 1 and 2 boiling water reactors

    SciTech Connect

    Diercks, D.R.

    1983-12-01

    Elbow assemblies and adjacent piping from the loops A and B core spray injection lines of Quad Cities Units 1 and 2 Boiling Water Reactors have been examined in order to determine the nature and causes of coolant leakages and flaw indications detected during hydrostatic tests and subsequent ultrasonic inspections. The elbow assemblies were found to contain multiple intergranular cracks in the weld heat-affected zones. The cracking was predominantly axial in orientation in the forged elbow and wedge components, whereas mixed axial and circumferential cracking was seen in the wrought piping pieces. In at least two instances, axial cracks completely penetrated the circumferential weld joining adjacent components. Based upon the observations made in the present study, the failures were attributed to intergranular stress corrosion cracking caused by the weld-induced sensitized microstructure and residual stresses present; dissolved oxygen in the reactor coolant apparently served as the corrosive species. The predominantly axial orientation of the cracks present in the forged components is believed to be related to the banded microstructure present in these components. The metallographic studies reported are supplemented by x-radiography, chemical analysis and mechanical test results, determinations of the degree of sensitization present, and measurements of weld metal delta ferrite content.

  17. Cold neutron tomography of annular coolant flow in a double subchannel model of a boiling water reactor

    NASA Astrophysics Data System (ADS)

    Kickhofel, J. L.; Zboray, R.; Damsohn, M.; Kaestner, A.; Lehmann, E. H.; Prasser, H.-M.

    2011-09-01

    Dryout of the liquid coolant film on fuel pins at the top of boiling water reactor (BWR) cores constitutes the type of heat transfer crisis relevant for the conditions of high void fractions. It is a limiting factor in the thermal power, and therefore the economy, of BWRs. Ongoing research on multiphase annular flow, specifically the liquid film thickness, is fundamental not only to nuclear reactor safety and operation but also to that of evaporators, condensers, and pipelines in a general industrial context. We have performed cold neutron tomography of adiabatic air water annular flow in a scaled up model of the subchannel geometry found in BWR fuel assemblies today. All imaging has been performed at the ICON beamline at the neutron spallation source SINQ at the Paul Scherrer Institut in Switzerland. Neutron tomography is shown to excel in investigating the interactions of air water two phase flows with spacer vanes of different geometry. The high resolution, high contrast measurements provide spatial distributions of the coolant on top of the surfaces of the spacer, including the vanes, and in the subchannel downstream of the spacers.

  18. Performance of boiling water reactor fuel lead test assemblies to 35 MWd/kg U

    SciTech Connect

    Rowland, T.C.; Ikemoto, R.N.; Gehl, S.

    1986-01-01

    This joint Electric Power Research Institute/General Electric (EPRI/GE) fuel performance program involved thorough preirradiation characterization of fuel used in lead test assemblies (LTAs), detailed surveillance of their operation, and interim site examinations of the assemblies during reactor outages. The program originally included four GE-5 LTAs operating in the Peach Bottom-2 (PB-2) reactor. The program was later modified to include the pressurized fuel rod test assembly in the Peach Bottom-3 (PB-3) reactor. The program modification also included extending the operation of the PB-2 and PB-3 LTA fuel beyond normal discharge exposures. Results are summarized in the paper.

  19. Feasibility assessment of burnup credit in the criticality analysis of shipping casks with boiling water reactor spent fuel

    SciTech Connect

    Broadhead, B.L.

    1991-08-01

    Considerable interest in the allowance of reactivity credit for the exposure history of power reactor fuel currently exists. This burnup credit'' issue has the potential to greatly reduce risk and cost when applied to the design and certification of spent fuel casks used for transportation and storage. Recently, analyses have demonstrated the technical feasibility and estimated the risk and economic incentives for allowing burnup credit in pressurized water reactor (PWR) spent fuel shipping cask applications. This report summarizes the extension of the previous PWR technical feasibility assessment to boiling water reactor (BWR) fuel. This feasibility analysis aims to apply simple methods that adequately characterize the time-dependent isotopic compositions of typical BWR fuel. An initial analysis objective was to identify a simple and reliable method for characterizing BWR spent fuel. Two different aspects of fuel characterization were considered:l first, the generation of burn- up dependent material interaction probabilities; second, the prediction of material inventories over time (depletion). After characterizing the spent fuel at various stages of exposure and decay, three dimensional (3-D) models for an infinite array of assemblies and, in several cases, infinite arrays of assemblies in a typical shipping cask basket were analyzed. Results for assemblies without a basket provide reactivity control requirements as a function of burnup and decay, while results including the basket allow assessment of typical basket configurations to provide sufficient reactivity control for spent BWR fuel. Resulting basket worths and reactivity trends over time are then evaluated to determine whether burnup credit is needed and feasible in BWR applications.

  20. Feasibility of underwater welding of highly irradiated in-vessel components of boiling-water reactors: A literature review

    SciTech Connect

    Lund, A.L.

    1997-11-01

    In February 1997, the U.S. Nuclear Regulatory Commission (NRC), Office of Nuclear Regulatory Research (RES), initiated a literature review to assess the state of underwater welding technology. In particular, the objective of this literature review was to evaluate the viability of underwater welding in-vessel components of boiling water reactor (BWR) in-vessel components, especially those components fabricated from stainless steels that are subjected to high neutron fluences. This assessment was requested because of the recent increased level of activity in the commercial nuclear industry to address generic issues concerning the reactor vessel and internals, especially those issues related to repair options. This literature review revealed a preponderance of general information about underwater welding technology, as a result of the active research in this field sponsored by the U.S. Navy and offshore oil and gas industry concerns. However, the literature search yielded only a limited amount of information about underwater welding of components in low-fluence areas of BWR in-vessel environments, and no information at all concerning underwater welding experiences in high-fluence environments. Research reported by the staff of the U.S. Department of Energy (DOE) Savannah River Site and researchers from the DOE fusion reactor program proved more fruitful. This research documented relevant experience concerning welding of stainless steel materials in air environments exposed to high neutron fluences. It also addressed problems with welding highly irradiated materials, and primarily attributed those problems to helium-induced cracking in the material. (Helium is produced from the neutron irradiation of boron, an impurity, and nickel.) The researchers found that the amount of helium-induced cracking could be controlled, or even eliminated, by reducing the heat input into the weld and applying a compressive stress perpendicular to the weld path.

  1. Final safety evaluation report related to the certification of the Advanced Boiling Water Reactor design. Supplement 1

    SciTech Connect

    1997-05-01

    This report supplements the final safety evaluation report (FSER) for the US Advanced Boiling Water Reactor (ABWR) standard design. The FSER was issued by the US Nuclear Regulatory Commission (NRC) staff as NUREG-1503 in July 1994 to document the NRC staff`s review of the US ABWR design. The US ABWR design was submitted by GE Nuclear Energy (GE) in accordance with the procedures of Subpart B to Part 52 of Title 10 of the Code of Federal Regulations. This supplement documents the NRC staff`s review of the changes to the US ABWR design documentation since the issuance of the FSER. GE made these changes primarily as a result of first-of-a-kind-engineering (FOAKE) and as a result of the design certification rulemaking for the ABWR design. On the basis of its evaluations, the NRC staff concludes that the confirmatory issues in NUREG-1503 are resolved, that the changes to the ABWR design documentation are acceptable, and that GE`s application for design certification meets the requirements of Subpart B to 10 CFR Part 52 that are applicable and technically relevant to the US ABWR design.

  2. Void fraction distribution in a boiling water reactor fuel assembly and the evaluation of subchannel analysis codes

    SciTech Connect

    Inoue, Akira; Futakuchi, Masanobu; Yagi, Makoto; Mitsutake, Toru; Morooka, Shinichi

    1995-12-01

    Void fraction measurement tests for boiling water reactor (BWR) simulated nuclear fuel assemblies have been conducted using an X-ray computed tomography scanner.there are two types of fuel assemblies concerning water rods. One fuel assembly has two water rods; the other has one large water rod. The effects of the water rods on radial void fraction distributions are measured within the fuel assemblies. The results show that the water rod effect does not make a large difference in void fraction distribution. The subchannel analysis codes COBRA/BWR and THERMIT-2 were compared with subchannel-averaged void fractions. The prediction accuracy of COBRA/BWR and THERMIT-2 for the subchannel-averaged void fraction was {Delta}{alpha} = {minus}3.6%, {sigma} = 4.8% and {Delta}{alpha} = {minus}4.1%, {sigma} = 4.5%, respectively, where {Delta}{alpha} is the average of the difference measured and calculated values. The subchannel analysis codes are highly applicable for the prediction of a two-phase flow distribution within BWR fuel assemblies.

  3. Prediction of dryout performance for boiling water reactor fuel assemblies based on subchannel analysis with the RINGS code

    SciTech Connect

    Knabe, P.; Wehle, F.

    1995-12-01

    A fuel assembly with a large critical power margin introduces flexibility into reload fuel management. Therefore, optimization of the bundle and spacer geometry to maximize the bundle critical power is an important design objective. With a view to reducing the extent of the complex full-scale tests usually carried out to determine the thermal-hydraulic characteristics of various assembly geometries, the subchannel analysis method was further developed with the Siemens RINGS code. The annular flow code predicts dryout power and dryout location by calculating the conditions at which the liquid film flow rate is reduced to zero, allowing for evaporation, droplet entrainment, and droplet deposition. Appropriate attention is paid to the modeling of spacer effects. Comparison with experimental data of 3 x 3 and 4 x 4 tests shows the capability of RINGS to predict the flow quality and mass flux in subchannels under typical boiling water reactor operating conditions. By using the RINGS code, experimental critical power data for 3 x 3, 4 x 4, 5 x 5, 7 x 7, 8 x 8, 9 x 9, and 10 x 10 fuel assemblies were successfully postcalculated.

  4. Experimental study of the effect of void reactivity feedback on the behavior of the scaled model boiling water reactor

    NASA Astrophysics Data System (ADS)

    Meftah, Khaled

    A Scaled Model Boiling Water Reactor (SMBWR) model uses low pressure (i.e., 0.095 MPa) water in a heated channel 0.5 meters in length with four electrically heated fuel simulator rods. The axial void profile in the channel is measured using conductivity probes and the power to the heaters is modulated according to the void fraction to simulate void reactivity feedback. The steam from the heated channel is passed through a valve that reduces the pressure to 0.012 MPa where the steam is condensed in conditions similar to those found in a conventional BWR condenser. The feedwater flow rate, heater power, and instrumentation in the facility are controlled and monitored through a Quadra 950 computer running LabVIEW software. The void fraction signals are analyzed to identify the different flow regimes and determine the vapor velocity in the SMBWR channel using features of the probability density function and power spectral density. The void coefficient of reactivity is modified in the BWR scale model through the LabVIEW interface and the effect on the behavior of the channel is directly observed. The system response is reported for abrupt stepwise pressure changes and abrupt stepwise power changes. The response is typical of that expected for a BWR. The void reactivity feedback effect is also examined by analyzing the frequency response of the channel void fraction at steady state.

  5. 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.

  6. Optimization of boiling water reactor control rod patterns using linear search

    SciTech Connect

    Kiguchi, T.; Doi, K.; Fikuzaki, T.; Frogner, B.; Lin, C.; Long, A.B.

    1984-10-01

    A computer program for searching the optimal control rod pattern has been developed. The program is able to find a control rod pattern where the resulting power distribution is optimal in the sense that it is the closest to the desired power distribution, and it satisfies all operational constraints. The search procedure consists of iterative uses of two steps: sensitivity analyses of local power and thermal margins using a three-dimensional reactor simulator for a simplified prediction model; linear search for the optimal control rod pattern with the simplified model. The optimal control rod pattern is found along the direction where the performance index gradient is the steepest. This program has been verified to find the optimal control rod pattern through simulations using operational data from the Oyster Creek Reactor.

  7. Thermal-hydraulic instabilities in pressure tube graphite - moderated boiling water reactors

    SciTech Connect

    Tsiklauri, G.; Schmitt, B.

    1995-09-01

    Thermally induced two-phase instabilities in non-uniformly heated boiling channels in RBMK-1000 reactor have been analyzed using RELAP5/MOD3 code. The RELAP5 model of a RBMK-1000 reactor was developed to investigate low flow in a distribution group header (DGH) supplying 44 fuel pressure tubes. The model was evaluated against experimental data. The results of the calculations indicate that the period of oscillation for the high power tube varied from 3.1s to 2.6s, over the power range of 2.0 MW to 3.0 MW, respectively. The amplitude of the flow oscillation for the high powered tube varied from +100% to -150% of the tube average flow. Reverse flow did not occur in the lower power tubes. The amplitude of oscillation in the subcooled region at the inlet to the fuel region is higher than in the saturated region at the outlet. In the upper fuel region and outlet connectors the flow oscillations are dissipated. The threshold of flow instability for the high powered tubes of a RBMK reactor is compared to Japanese data and appears to be in good agreement.

  8. Characterization of Phases in “Crud” from Boiling Water Reactors by Transmission Electron Microscopy

    SciTech Connect

    Dawn E. Janney; Douglas L. Porter

    2007-05-01

    This paper reports phases identified in samples of crud (activated corrosion products) from two commercial boilingwater reactors using transmission and analytical electron microscopy and selected-area electron diffraction. Franklinite (ZnFe2O4) was observed in both samples. Hematite (a-Fe2O3), crystalline silica (SiO2), a fine-grained mixture of iron oxides probably including magnetite (Fe3O4), hematite (a-Fe2O3), and goethite (a-FeOOH), and an unidentified high-Ba, high-S phase were observed in one of the samples. Willemite (Zn2SiO4), amorphous silica, and an unidentified iron– chromium phase were observed in the other. Chloride-bearing phases were found in both samples, and are assumed to represent sample contaminants. Because of the small sample volumes and numbers of particles studied and the possibility of contamination, it is not clear whether the differences between the phases observed in the two crud samples represent actual differences in the assemblages formed in the reactors.

  9. Lifetime Neutron Fluence Analysis of the Ringhals Unit 1 Boiling Water Reactor

    NASA Astrophysics Data System (ADS)

    Kulesza, Joel A.; Roudén, Jenny; Green, Eva-Lena

    2016-02-01

    This paper describes a neutron fluence assessment considering the entire commercial operating history (35 cycles or ˜ 25 effective full power years) of the Ringhals Unit 1 reactor pressure vessel beltline region. In this assessment, neutron (E >1.0 MeV) fluence and iron atom displacement distributions were calculated on the moderator tank and reactor pressure vessel structures. To validate those calculations, five in-vessel surveillance chain dosimetry sets were evaluated as well as material samples taken from the upper core grid and wide range neutron monitor tubes to act as a form of retrospective dosimetry. During the analysis, it was recognized that delays in characterizing the retrospective dosimetry samples reduced the amount of reactions available to be counted and complicated the material composition determination. However, the comparisons between the surveillance chain dosimetry measurements (M) and calculated (C) results show similar and consistent results with the linear average M/C ratio of 1.13 which is in good agreement with the resultant least squares best estimate (BE)/C ratios of 1.10 for both neutron (E >1.0 MeV) flux and iron atom displacement rate.

  10. Chimney for enhancing flow of coolant water in natural circulation boiling water reactor

    DOEpatents

    Oosterkamp, Willem Jan; Marquino, Wayne

    1999-01-05

    A chimney which can be reconfigured or removed during refueling to allow vertical removal of the fuel assemblies. The chimney is designed to be collapsed or dismantled. Collapse or dismantlement of the chimney reduces the volume required for chimney storage during the refueling operation. Alternatively, the chimney has movable parts which allow reconfiguration of its structure. In a first configuration suitable for normal reactor operation, the chimney is radially constricted such that the chimney obstructs vertical removal of the fuel assemblies. In a second configuration suitable for refueling or maintenance of the fuel core, the parts of the chimney which obstruct access to the fuel assemblies are moved radially outward to positions whereat access to the fuel assemblies is not obstructed.

  11. Chimney for enhancing flow of coolant water in natural circulation boiling water reactor

    DOEpatents

    Oosterkamp, W.J.; Marquino, W.

    1999-01-05

    A chimney which can be reconfigured or removed during refueling to allow vertical removal of the fuel assemblies is disclosed. The chimney is designed to be collapsed or dismantled. Collapse or dismantlement of the chimney reduces the volume required for chimney storage during the refueling operation. Alternatively, the chimney has movable parts which allow reconfiguration of its structure. In a first configuration suitable for normal reactor operation, the chimney is radially constricted such that the chimney obstructs vertical removal of the fuel assemblies. In a second configuration suitable for refueling or maintenance of the fuel core, the parts of the chimney which obstruct access to the fuel assemblies are moved radially outward to positions whereas access to the fuel assemblies is not obstructed. 11 figs.

  12. Corrosion product deposits on boiling-water reactor cladding: Experimental and theoretical investigation of magnetic properties

    NASA Astrophysics Data System (ADS)

    Orlov, A.; Degueldre, C.; Wiese, H.; Ledergerber, G.; Valizadeh, S.

    2011-09-01

    Recent Eddy current investigations on the cladding of nuclear fuel pins have shown that the apparent oxide layers are falsified due to unexpected magnetic properties of corrosion product deposits. Analyses by Scanning Electron Microscopy (SEM) or Electron Probe Micro Analysis (EPMA) demonstrated that the deposit layer consists of complex 3-d element oxides (Ni, Mn, Fe) along with Zn, since the reactor operates with a Zn addition procedure to reduce buildup of radiation fields on the recirculation system surfaces. The oxides crystallise in ferritic spinel structures. These spinels are well-known for their magnetic behaviour. Since non-magnetic zinc ferrite (ZnFe 2O 4) may become magnetic when doped with even small amounts of Ni and/or Mn, their occurrence in the deposit layer has been analyzed. The magnetic permeability of zinc ferrite, trevorite and jacobsite and their solid solutions are estimated by magnetic moment additivity. From the void history examination, the low elevation sample (810 mm) did not face significant boiling during the irradiation cycles suggesting growth of (Mn0.092+Zn0.752+Fe0.293+)[(Fe1.713+Mn0.032+Ni0.132+)O] crystals with theoretical value of the magnetic permeability for the averaged heterogeneous CRUD layer of 9.5 ± 3. Meanwhile, (Mn0.162+Zn0.552+Fe0.293+)[(Fe1.713+Mn0.042+Ni0.252+)O] crystallizes at the mid elevation (1810 mm) with theoretical magnetic permeability for the CRUD layer of 4.2 ± 1.5 at the investigated azimuthal location. These theoretical data are compared with the magnetic permeability of the corrosion product deposited layers gained from reactor pool side Eddy current (EC) analyses (9.0 ± 1.0 for low and 3.5 ± 1.0 for high elevation). The calculated thicknesses and magnetic permeability values of the deposition layers (estimated by MAGNACROX multifrequency EC method) match together with these estimated using an "ion magnetic moment additivity" model.

  13. Bottom head to shell junction assembly for a boiling water nuclear reactor

    DOEpatents

    Fife, Alex Blair; Ballas, Gary J.

    1998-01-01

    A bottom head to shell junction assembly which, in one embodiment, includes an annular forging having an integrally formed pump deck and shroud support is described. In the one embodiment, the annular forging also includes a top, cylindrical shaped end configured to be welded to one end of the pressure vessel cylindrical shell and a bottom, conical shaped end configured to be welded to the disk shaped bottom head. Reactor internal pump nozzles also are integrally formed in the annular forging. The nozzles do not include any internal or external projections. Stubs are formed in each nozzle opening to facilitate welding a pump housing to the forging. Also, an upper portion of each nozzle opening is configured to receive a portion of a diffuser coupled to a pump shaft which extends through the nozzle opening. Diffuser openings are formed in the integral pump deck to provide additional support for the pump impellers. The diffuser opening is sized so that a pump impeller can extend at least partially therethrough. The pump impeller is connected to the pump shaft which extends through the nozzle opening.

  14. Bottom head to shell junction assembly for a boiling water nuclear reactor

    DOEpatents

    Fife, A.B.; Ballas, G.J.

    1998-02-24

    A bottom head to shell junction assembly which, in one embodiment, includes an annular forging having an integrally formed pump deck and shroud support is described. In the one embodiment, the annular forging also includes a top, cylindrical shaped end configured to be welded to one end of the pressure vessel cylindrical shell and a bottom, conical shaped end configured to be welded to the disk shaped bottom head. Reactor internal pump nozzles also are integrally formed in the annular forging. The nozzles do not include any internal or external projections. Stubs are formed in each nozzle opening to facilitate welding a pump housing to the forging. Also, an upper portion of each nozzle opening is configured to receive a portion of a diffuser coupled to a pump shaft which extends through the nozzle opening. Diffuser openings are formed in the integral pump deck to provide additional support for the pump impellers. The diffuser opening is sized so that a pump impeller can extend at least partially therethrough. The pump impeller is connected to the pump shaft which extends through the nozzle opening. 5 figs.

  15. Preliminary Feasibility, Design, and Hazard Analysis of a Boiling Water Test Loop Within the Idaho National Laboratory Advanced Test Reactor National Scientific User Facility

    SciTech Connect

    Douglas M. Gerstner

    2009-05-01

    The Advanced Test Reactor (ATR) is a pressurized light-water reactor with a design thermal power of 250 MW. The principal function of the ATR is to provide a high neutron flux for testing reactor fuels and other materials. The ATR and its support facilities are located at the Idaho National Laboratory (INL). A Boiling Water Test Loop (BWTL) is being designed for one of the irradiation test positions within the. The objective of the new loop will be to simulate boiling water reactor (BWR) conditions to support clad corrosion and related reactor material testing. Further it will accommodate power ramping tests of candidate high burn-up fuels and fuel pins/rods for the commercial BWR utilities. The BWTL will be much like the pressurized water loops already in service in 5 of the 9 “flux traps” (region of enhanced neutron flux) in the ATR. The loop coolant will be isolated from the primary coolant system so that the loop’s temperature, pressure, flow rate, and water chemistry can be independently controlled. This paper presents the proposed general design of the in-core and auxiliary BWTL systems; the preliminary results of the neutronics and thermal hydraulics analyses; and the preliminary hazard analysis for safe normal and transient BWTL and ATR operation.

  16. Technology, safety and costs of decommissioning a reference boiling water reactor power station. Volume 2. Appendices. Technical report, September 1977-October 1979

    SciTech Connect

    Oak, H.D.; Holter, G.M.; Kennedy, W.E. Jr.; Konzek, G.J.

    1980-06-01

    Technology, safety and cost information is given for the conceptual decommissioning of a large (1100MWe) boiling water reactor (BWR) power station. Three approaches to decommissioning, immediate dismantlement, safe storage with deferred dismantlement and entombment, were studied to obtain comparisons between costs, occupational radiation doses, potential dose to the public and other safety impacts. It also shows the sensitivity of decommissioning safety and costs to the power rating of a BWR in the range of 200 to 1100 MWE. This volume contains the appendices.

  17. Pulsational characteristics of the natural-circulation loop of a large-scale model of a light-boiling boiling-water reactor

    SciTech Connect

    Babykin, A.S.; Balunov, B.F.; Chernykh, N.G.; Smirnov, E.L.; Tisheninova, V.I.; Zhiuitskaya, T.S.

    1985-10-01

    The results of an experimental study of a natural-circulation (NC) loop, whose geometrical and hydraulic characteristics are presented are described. The range of state parameters encompassed in the experiments is also indicated. The authors used a large-scale model of a low-boiling-water reactor, with natural heights and reduced stages of separate elements of the NC loop. The study confirmed that under the conditions the pulsations in the flow rate of the coolant occurs only in the transitional zone from natural circulation of the singlephase medium to natural circulation of the two-phase coolant.

  18. Comparison of Standard Light Water Reactor Cross-Section Libraries using the United States Nuclear Regulatory Commission Boiling Water Reactor Benchmark Problem

    NASA Astrophysics Data System (ADS)

    Kulesza, Joel A.; Arzu Alpan, F.

    2016-02-01

    This paper describes a comparison of contemporary and historical light water reactor shielding and pressure vessel dosimetry cross-section libraries for a boiling water reactor calculational benchmark problem. The calculational benchmark problem was developed at Brookhaven National Laboratory by the request of the U. S. Nuclear Regulatory Commission. The benchmark problem was originally evaluated by Brookhaven National Laboratory using the Oak Ridge National Laboratory discrete ordinates code DORT and the BUGLE-93 cross-section library. In this paper, the Westinghouse RAPTOR-M3G three-dimensional discrete ordinates code was used. A variety of cross-section libraries were used with RAPTOR-M3G including the BUGLE93, BUGLE-96, and BUGLE-B7 cross-section libraries developed at Oak Ridge National Laboratory and ALPAN-VII.0 developed at Westinghouse. In comparing the calculated fast reaction rates using the four aforementioned cross-section libraries in the pressure vessel capsule, for six dosimetry reaction rates, a maximum relative difference of 8% was observed. As such, it is concluded that the results calculated by RAPTOR-M3G are consistent with the benchmark and further that the different vintage BUGLE cross-section libraries investigated are largely self-consistent.

  19. Experimental study of the interaction of pulsations of the neutron flux and the coolant flow in a boiling-water reactor

    SciTech Connect

    Leppik, P.A.

    1984-12-01

    This paper presents results of a study designed to confirm that the interaction of the neutron flux and the coolant flow plays an important role in the mechanism of high-frequency (HF) resonant instability of the VK-50 boiling water reactor. To do this and to check the working model, signals from probes measuring the flow rate of the coolant and the neutron flux were recorded simultaneously (with the help of a magnetograph) in experiments performed in 1981 on driving the VK-50 reactor into the HF reonant instability regimes. Estimates were then obtained for the statistical characteristics of the pulsations of the flow rate and of the neutron flux, including the cross-correlation functions and coherence functions. The basic results of these studies are reported here.

  20. Light Water Reactor Sustainability Program Support and Modeling for the Boiling Water Reactor Station Black Out Case Study Using RELAP and RAVEN

    SciTech Connect

    Diego Mandelli; Curtis Smith; Thomas Riley; John Schroeder; Cristian Rabiti; Aldrea Alfonsi; Joe Nielsen; Dan Maljovec; Bie Wang; Valerio Pascucci

    2013-09-01

    The existing fleet of nuclear power plants is in the process of extending its lifetime and increasing the power generated. In order to evaluate the impact of these two 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 impact of power uprate on the safety of a boiled water reactor system. The case study considered is a loss of off-site power followed by the loss of diesel generators, i.e., a station black out (SBO) event. Analysis is performed by using a thermo-hydraulic code, i.e. RELAP-5, and a stochastic analysis tool currently under development at INL, i.e. RAVEN. Starting from the event tree models contained in SAPHIRE, we built the input file for RELAP-5 that models in great detail system dynamics under SBO conditions. We also interfaced RAVEN with RELAP-5 so that it would be possible to run multiple RELAP-5 simulation runs by changing specific keywords of the input file. We both employed classical statistical tools, i.e. Monte-Carlo, and more advanced machine learning based algorithms to perform uncertainty quantification in order to quantify changes in system performance and limitations as a consequence of power uprate. We also employed advanced data analysis and visualization tools that helped us to correlate simulation outcome such as maximum core temperature with a set of input uncertain parameters. Results obtained gave a detailed overview of the issues associated to power uprate for a SBO accident scenario. We were able to quantify how timing of safety related events were impacted by a higher reactor core power. Such insights can provide useful material to the decision makers to perform risk-infomed safety margins management.

  1. Estimating boiling water reactor decommissioning costs: A user`s manual for the BWR Cost Estimating Computer Program (CECP) software. Final report

    SciTech Connect

    Bierschbach, M.C.

    1996-06-01

    Nuclear power plant licensees are required to submit to the US Nuclear Regulatory Commission (NRC) for review their decommissioning cost estimates. This user`s manual and the accompanying Cost Estimating Computer Program (CECP) software provide a cost-calculating methodology to the NRC staff that will assist them in assessing the adequacy of the licensee submittals. The CECP, designed to be used on a personal computer, provides estimates for the cost of decommissioning boiling water reactor (BWR) power stations to the point of license termination. Such cost estimates include component, piping, and equipment removal costs; packaging costs; decontamination costs; transportation costs; burial costs; and manpower costs. In addition to costs, the CECP also calculates burial volumes, person-hours, crew-hours, and exposure person-hours associated with decommissioning.

  2. Radiolysis of boiling water

    NASA Astrophysics Data System (ADS)

    Yang, Shuang; Katsumura, Yosuke; Yamashita, Shinichi; Matsuura, Chihiro; Hiroishi, Daisuke; Lertnaisat, Phantira; Taguchi, Mitsumasa

    2016-06-01

    γ-radiolysis of boiling water has been investigated. The G-value of H2 evolution was found to be very sensitive to the purity of water. In high-purity water, both H2 and O2 gases were formed in the stoichiometric ratio of 2:1; a negligible amount of H2O2 remained in the liquid phase. The G-values of H2 and O2 gas evolution depend on the dose rate: lower dose rates produce larger yields. To clarify the importance of the interface between liquid and gas phase for gas evolution, the gas evolution under Ar gas bubbling was measured. A large amount of H2 was detected, similar to the radiolysis of boiling water. The evolution of gas was enhanced in a 0.5 M NaCl aqueous solution. Deterministic chemical kinetics simulation elucidated the mechanism of radiolysis in boiling water.

  3. Superfund record of decision (EPA region 10): Idaho National Engineering Lab, (USDOE) Operable Unit 26 (Stationary Low-Power Reactor-1 and Boiling Water Reactor Experiment-I Burial Grounds), Idaho Falls, ID, December 1, 1995

    SciTech Connect

    1997-03-01

    This document presents the selected remedial action for the Stationary Low-Power Reactor-1 (SL-1) burial ground, the Boiling Water Reactor Experiment-I (BORAX-I) burial ground, and 10 no action sites in Waste Area Group 5. Actual or threatened releases of hazardous substances from the SL-1 and BORAX-I burial grounds, if not addressed by implementing the response action selected in this Record of Decision, may present a current or potential threat to public health, welfare, or the environment. The 10 no action sites do not present a threat to human health or the environment.

  4. On the shape of stress corrosion cracks in sensitized Type 304 SS in Boiling Water Reactor primary coolant piping at 288 °C

    NASA Astrophysics Data System (ADS)

    Lee, Sang-Kwon; Kramer, Daniel; Macdonald, Digby D.

    2014-11-01

    Evolution of the shape of surface cracks in sensitized Type 304 SS in Boiling Water Reactor primary coolant circuit piping at the reactor operating temperature of 288 °C is explored as a function of various environmental variables, such as electrochemical potential (ECP), solution conductivity, flow velocity, and multiplier for the oxygen reduction reaction (ORR) standard exchange current density (SECD), using the coupled environment fracture model (CEFM). For this work, the CEFM was upgraded by incorporating Shoji's model for calculating the crack tip strain rate and more advanced expressions were used for estimating the stress intensity factor for semi-elliptical surface cracks. This revised CEFM accurately predicts the dependence of the crack growth rate on stress intensity factor and offers an alternative explanation for the development of semi-elliptical cracks than that provided by fracture mechanics alone. The evolution of surface crack semi-elliptical shape depends strongly upon various environmental variables identified above, and the CEFM predicts that the minor axis of the ellipse should be oriented perpendicular to the surface, in agreement with observation. The development of the observed semi-elliptical cracks with the minor axis perpendicular to the surface is therefore attributed to the dependence of the crack growth rate on the electrochemical crack length.

  5. Spent fuel assembly hardware: Characterization and 10 CFR 61 classification for waste disposal: Volume 3, Calculated activity profiles of spent nuclear fuel assembly hardware for boiling water reactors

    SciTech Connect

    Short, S.M.; Luksic, A.T.; Schutz, M.E.

    1989-06-01

    Consolidation of spent fuel is under active consideration as the US Department of Energy plans to dispose of spent fuel as required by the Nuclear Waste Policy Act of 1982. During consolidation, the fuel pins are removed from an intact fuel assembly and repackaged into a more compact configuration. After repackaging, approximately 30 kg of residual spent fuel assembly hardware per assembly that is also radioactive and required disposal. Understanding the nature of this secondary waste stream is critical to designing a system that will properly handle, package, store, and dispose of the waste. This report presents a methodology for estimating the radionuclide inventory in irradiated spent fuel hardware. Ratios are developed that allow the use of ORIGEN2 computer code calculations to be applied to regions that are outside the fueled region. The ratios are based on the analysis of samples of irradiated hardware from spent fuel assemblies. The results of this research are presented in three volumes. In Volume 1, the development of scaling factors that can be used with ORIGEN2 calculations to estimate activation of spent fuel assembly hardware is documented. The results from laboratory analysis of irradiated spent-fuel hardware samples are also presented in Volume 1. In Volume 2 and 3, the calculated flux profiles of spent nuclear fuel assemblies are presented for pressurized water reactors and boiling water reactors, respectively. The results presented in Volumes 2 and 3 were used to develop the scaling factors documented in Volume 1.

  6. 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.

  7. The Results of Feasibility Study of Co-generation NPP With Innovative VK-300 Simplified Boiling Water Reactor

    SciTech Connect

    Kuznetsov, Yury N.

    2006-07-01

    The co-generation nuclear power plant (CNPP) producing electricity and district heating heat is planned to be constructed in Archangelsk Region of Russia. Following the 'Letter of Intent' signed by Governor of Archangelsk region and by Minister of the Russian Federation for atomic energy the feasibility study of the Project has been done. The NPP will be based on the four co-generation nuclear power units with the Russian VK-300 SBWR. The innovative passive VK-300 reactor facility has been designed on the basis of well-established nuclear technologies, proven major components, the operating experience of the prototype VK-50 reactor in RIAR, Dimitrovgrad, and the experience in designing such reactors as SBWR (GE) and SWR-1000 (Siemens). The CNPP's total power is planned to be 1000 MW(e) and district-heating heat production capacity 1600 Gcal/h. A detailed description of the results of the feasibility study is presented in the report. The results of the feasibility study have shown that the Archangelsk CGNP is feasible in terms of engineering, economics and production. (authors)

  8. The low-power low-pressure flow resonance in a natural circulation cooled boiling water reactor

    SciTech Connect

    Hagen, T.H.J.J. van der; Stekelenburg, A.J.C.

    1995-09-01

    The last few years the possibility of flow resonances during the start-up phase of natural circulation cooled BWRs has been put forward by several authors. The present paper reports on actual oscillations observed at the Dodewaard reactor, the world`s only operating BWR cooled by natural circulation. In addition, results of a parameter study performed by means of a simple theoretical model are presented. The influence of relevant parameters on the resonance characteristics, being the decay ratio and the resonance frequency, is investigated and explained.

  9. E-chem page: A Support System for Remote Diagnosis of Water Quality in Boiling Water Reactors

    SciTech Connect

    Naohiro Kusumi; Takayasu Kasahara; Kazuhiko Akamine; Kenji Tada; Naoshi Usui; Nobuyuki Oota

    2002-07-01

    It is important to control and maintain water quality for nuclear power plants. Chemical engineers sample and monitor reactor water from various subsystems and analyze the chemical quality as routine operations. With regard to controlling water quality, new technologies have been developed and introduced to improve the water quality from both operation and material viewpoints. To maintain the quality, it is important to support chemical engineers in evaluating the water quality and realizing effective retrieval of stored data and documents. We have developed a remote support system using the Internet to diagnose BWR water quality, which we call e-chem page. The e-chem page integrates distributed data and information in a Web server, and makes it easy to evaluate the data on BWR water chemistry. This system is composed of four functions: data transmission, water quality evaluation, inquiry and history retrieval system, and reference to documents on BWR water chemistry. The developed system is now being evaluated in trial operations by Hitachi, Ltd. and an electric power company. In addition diagnosis technology applying independent component analysis (ICA) is being developed to improve predictive capability of the system. This paper describes the structure and function of the e-chem page and presents results of obtained with the proposed system for the prediction of chemistry conditions in reactor water. (authors)

  10. An assessment of BWR (boiling water reactor) Mark-II containment challenges, failure modes, and potential improvements in performance

    SciTech Connect

    Kelly, D.L.; Jones, K.R.; Dallman, R.J. ); Wagner, K.C. )

    1990-07-01

    This report assesses challenges to BWR Mark II containment integrity that could potentially arise from severe accidents. Also assessed are some potential improvements that could prevent core damage or containment failure, or could mitigate the consequences of such failure by reducing the release of fission products to the environment. These challenges and improvements are analyzed via a limited quantitative risk/benefit analysis of a generic BWR/4 reactor with Mark II containment. Point estimate frequencies of the dominant core damage sequences are obtained and simple containment event trees are constructed to evaluate the response of the containment to these severe accident sequences. The resulting containment release modes are then binned into source term release categories, which provide inputs to the consequence analysis. The output of the consequences analysis is used to construct an overall base case risk profile. Potential improvements and sensitivities are evaluated by modifying the event tree spilt fractions, thus generating a revised risk profile. Several important sensitivity cases are examined to evaluate the impact of phenomenological uncertainties on the final results. 75 refs., 25 figs., 65 tabs.

  11. Use of an influence diagram and fuzzy probability for evaluating accident management in a boiling water reactor

    SciTech Connect

    Yu, D.; Kastenberg, W.E.; Okrent, D. . Mechanical, Aerospace, and Nuclear Engineering Dept.)

    1994-06-01

    A new approach is presented for evaluating the uncertainties inherent in severe accident management strategies. At first, this analysis considers accident management as a decision problem (i.e., applying a strategy compared with do nothing) and uses an influence diagram. To evaluate imprecise node probabilities in the influence diagram, the analysis introduces the concept of a fuzzy probability. When fuzzy logic is applied, fuzzy probabilities are easily propagated to obtain results. In addition, the results obtained provide not only information similar to the classical approach, which uses point-estimate values, but also additional information regarding the impact of using imprecise input data. As an illustrative example, the proposed methodology is applied to the evaluation of the drywell flooding strategy for a long-term station blackout sequence at the Peach Bottom nuclear power plant. The results show that the drywell flooding strategy is beneficial for preventing reactor vessel breach. It is also effective for reducing the probability of containment failure for both liner melt-through and late overpressurization. Even though uncertainty exists in the results, flooding is preferred to do nothing when evaluated in terms of two risk measures: early and late fatalities.

  12. A citation-based assessment of the performance of U.S. boiling water reactors following extended power up-rates

    NASA Astrophysics Data System (ADS)

    Heidrich, Brenden J.

    Nuclear power plants produce 20 percent of the electricity generated in the U.S. Nuclear generated electricity is increasingly valuable to a utility because it can be produced at a low marginal cost and it does not release any carbon dioxide. It can also be a hedge against uncertain fossil fuel prices. The construction of new nuclear power plants in the U.S. is cautiously moving forward, restrained by high capital costs. Since 1998, nuclear utilities have been increasing the power output of their reactors by implementing extended power up-rates. Power increases of up to 20 percent are allowed under this process. The equivalent of nine large power plants has been added via extended power up-rates. These up-rates require the replacement of large capital equipment and are often performed in concert with other plant life extension activities such as license renewals. This dissertation examines the effect of these extended power up-rates on the safety performance of U.S. boiling water reactors. Licensing event reports are submitted by the utilities to the Nuclear Regulatory Commission, the federal nuclear regulator, for a wide range of abnormal events. Two methods are used to examine the effect of extended power up-rates on the frequency of abnormal events at the reactors. The Crow/AMSAA model, a univariate technique is used to determine if the implementation of an extended power up-rate affects the rate of abnormal events. The method has a long history in the aerospace industry and in the military. At a 95-percent confidence level, the rate of events requiring the submission of a licensing event report decreases following the implementation of an extended power up-rate. It is hypothesized that the improvement in performance is tied to the equipment replacement and refurbishment that is performed as part of the up-rate process. The reactor performance is also analyzed using the proportional hazards model. This technique allows for the estimation of the effects of

  13. BWRSAR (Boiling Water Reactor Severe Accident Response) calculations of reactor vessel debris pours for Peach Bottom short-term station blackout

    SciTech Connect

    Hodge, S.A.; Ott, L.J.

    1988-01-01

    This paper describes recent analyses performed by the BWR Severe Accident Technology (BWRSAT) Program at Oak Ridge National Laboratory to estimate the release of debris from the reactor vessel for the unmitigated short-term station blackout accident sequence. Calculations were performed with the BWR Severe Accident Response (BWRSAR) code and are based upon consideration of the Peach Bottom Atomic Power Station. The modeling strategies employed within BWRSAR for debris relocation within the reactor vessel are briefly discussed and the calculated events of the accident sequence, including details of the calculated debris pours, are presented. 4 refs., 13 figs., 3 tabs.

  14. Study of Pu consumption in light water reactors: Evaluation of GE advanced boiling water reactor plants, compilation of Phase 1C task reports

    SciTech Connect

    Not Available

    1994-01-15

    This report summarizes the evaluations conducted during Phase 1C of the Pu Disposition Study have provided further results which reinforce the conclusions reached during Phase 1A & 1B: These conclusions clearly establish the benefits of the fission option and the use of the ABWR as a reliable, proven, well-defined and cost-effective means available to disposition the weapons Pu. This project could be implemented in the near-term at a cost and on a schedule being validated by reactor plants currently under construction in Japan and by cost and schedule history and validated plans for MOX plants in Europe. Evaluations conducted during this phase have established that (1) the MOX fuel is licensable based on existing criteria for new fuel with limited lead fuel rod testing, (2) that the applicable requirements for transport, handling and repository storage can be met, and (3) that all the applicable safeguards criteria can be met.

  15. Experimental study of void behavior in a suppression pool of a boiling water reactor during the blowdown period of a loss of coolant accident

    NASA Astrophysics Data System (ADS)

    Rassame, Somboon

    The possible failure of an Emergency Core Cooling System (ECCS) train due to a large amount of entrained gas in the ECCS pump suction piping in a Loss of Coolant Accident (LOCA) is one of the potential engineering problems faced in a Boiling Water Reactor (BWR) power plant. To analyze potential gas intrusion into the ECCS pump suction piping, the study of void behavior in the Suppression Pool (SP) during the LOCA is necessary. The void fraction distribution and void penetration are considered as the key parameters in the problem analysis. Two sets of experiments, namely, steady-state tests and transient tests were conducted using the Purdue University Multi-Dimensional Integral Test Assembly for ESBWR application (PUMA-E) to study void behavior in the SP during the blowdown. The design of the test apparatus used is based on the scaling analysis from a prototypical BWR containment (MARK-I) with consideration of the downcomer size, the SP water level, and the downcomer water submergence depth. Several instruments were installed to obtain the required experimental data, such as inlet gas volumetric flow, void fraction, pressure, and temperature. For the steady-state tests, the air was injected through a downcomer pipe in the SP in order to simulate the physical phenomena in the SP during the initial blowdown of LOCA. Thirty tests were performed with two different downcomer sizes (0.076 and 0.102 m), various air volumetric flow rates or flux (0.003 to 0.153 m3/s or 0.5 to 24.7 m/s), initial downcomer void conditions (fully filled with water, partially void, and completely void) and air velocity ramp rates (one to two seconds). Two phases of the experiment were observed, namely, the initial phase and the quasi-steady phase. The initial phase produced the maximum void penetration depth; and the quasi-steady phase showed less void penetration with oscillation in the void penetration. The air volumetric flow rate was found to have a minor effect on the void fraction

  16. Equalization of energy density in boiling water reactors (as exemplified by WB-50). Development and testing of WB -50 computational model on the basis of MCU-RR code

    NASA Astrophysics Data System (ADS)

    Chertkov, Yu B.; Disyuk, V. V.; Pimenov, E. Yu; Aksenova, N. V.

    2017-01-01

    Within the framework of research in possibility and prospects of power density equalization in boiling water reactors (as exemplified by WB-50) a work was undertaken to improve prior computational model of the WB-50 reactor implemented in MCU-RR software. Analysis of prior works showed that critical state calculations have deviation of calculated reactivity exceeding ±0.3 % (ΔKef/Kef) for minimum concentrations of boric acid in the reactor water and reaching 2 % for maximum concentration values. Axial coefficient of nonuniform burnup distribution reaches high values in the WB-50 reactor. Thus, the computational model needed refinement to take into account burnup inhomogeneity along the fuel assembly height. At this stage, computational results with mean square deviation of less than 0.7 % (ΔKef/Kef) and dispersion of design values of ±1 % (ΔK/K) shall be deemed acceptable. Further lowering of these parameters apparently requires root cause analysis of such large values and paying more attention to experimental measurement techniques.

  17. Environmentally-assisted cracking behaviour in the transition region of an Alloy182/SA 508 Cl.2 dissimilar metal weld joint in simulated boiling water reactor normal water chemistry environment

    NASA Astrophysics Data System (ADS)

    Seifert, H. P.; Ritter, S.; Shoji, T.; Peng, Q. J.; Takeda, Y.; Lu, Z. P.

    2008-08-01

    The stress corrosion cracking (SCC) and corrosion fatigue behaviour perpendicular and parallel to the fusion line in the transition region between the Alloy 182 Nickel-base weld metal and the adjacent SA 508 Cl.2 low-alloy reactor pressure vessel (RPV) steel of a simulated dissimilar metal weld joint was investigated under boiling water reactor normal water chemistry conditions. A special emphasis was placed to the question whether a fast growing interdendritic SCC crack in the highly susceptible Alloy 182 weld metal can easily cross the fusion line and significantly propagate into the adjacent low-alloy RPV steel. Cessation of interdendritic SCC crack growth was observed in high-purity or sulphate-containing oxygenated water under constant or periodical partial unloading conditions for those parts of the crack front, which reached the fusion line. In chloride containing water, on the other hand, the interdendritic SCC crack in the Alloy 182 weld metal very easily crossed the fusion line and further propagated with a very high rate as a transgranular crack into the heat-affected zone and base metal of the adjacent low-alloy steel. The observed SCC cracking behaviour at the interface correlates excellently with the field experience of such dissimilar metal weld joints, where SCC cracking was usually confined to the Alloy 182 weld metal.

  18. Review and evaluation of the RELAP5YA computer code and the Vermont Yankee LOCA (Loss-of-Coolant Accident) licensing analysis model for use in small and large break BWR (Boiling Water Reactor) LOCAS

    SciTech Connect

    Jones, J.L.

    1987-01-01

    A review has been completed of the RELAP5YA computer code to determine its acceptability for performing licensing analyses. The review was limited to Boiling Water Reactor (BWR) reactor applications. In addition, a Loss-Of-Coolant Accident (LOCA) licensing analysis method, using the RELAP5YA computer code, has been reviewed. This method is applicable to the Vermont Yankee Nuclear Power Station to perform full break spectra LOCA and fuel cycle independent analyses. The review of the RELAP5YA code consisted of an evaluation of all Yankee Atomic Electric Company (YAEC) incorporated modifications to the RELAP5/MOD1 Cycle 18 computer code from which the licensing version of the code originated. Qualifying separate and integral effects assessment calculations were reviewed to evaluate the validity and proper implementation of the various added models. The LOCA licensing method was assessed by reviewing two RELAP5YA system input models and evaluating several small and large break qualifying transient calculations. A review of the RELAP5YA code modifications and their assessments, as well as the submitted LOCA licensing method, is given and the results of the review are provided.

  19. Experimental Study on Thermal-Hydraulics During Start-Up in the Natural Circulation Boiling Water Reactor Concept: Effects of System Pressure and Increasing Heat Flux on the Geysering and Density Wave Oscillation

    SciTech Connect

    Hadid Subki, M.; Masanori Aritomi; Noriyuki Watanabe; Chaiwat Muncharoen

    2002-07-01

    The feasibility study in thermal-hydraulics for the future light water reactor concept is carried out. One of the essential studies is the two-phase flow instability during start-up in the natural circulation boiling water reactor (BWR) concept. It is anticipated that the occurrence of the two-phase flow instabilities during start-up significantly affects the feasibility concept, since it would cause the complexity in raising and maneuvering the power output. The purpose of the current study is to experimentally investigate the driving mechanism of the geysering and density wave oscillation in the natural circulation loop, induced by a range of system operating pressure and increasing heat flux in vertical parallel channels. The pressure range of atmospheric up to about 4 bars, and the input heat flux range of 0 up to 577 kW/m{sup 2} are applied in these experiments. An experimental apparatus of twin boiling upflow channels to simulate natural circulation flow loop has been designed, constructed and operated. The natural circulation in the loop occurs due to the density difference between two-phase region in the channels and the single-phase liquid in the downcomer. The objective of the study is to propose a rational start-up procedure in which the geysering and density wave oscillation can be prevented during startup, according to its system pressure and heat flux. Previous studies have clarified that three (3) kinds of thermo-hydraulics instabilities may occur during start-up in the natural circulation BWR depending on its procedure and reactor configuration, which are (1) geysering induced by condensation, (2) natural circulation induced by hydrostatic head fluctuation in steam separator, and (3) density wave oscillation. (authors)

  20. Revised analyses of decommissioning for the reference boiling water reactor power station. Effects of current regulatory and other considerations on the financial assurance requirements of the decommissioning rule and on estimates of occupational radiation exposure - main report. Final report

    SciTech Connect

    Smith, R.I.; Bierschbach, M.C.; Konzek, G.J.; McDuffie, P.N.

    1996-07-01

    The NRC staff is in need of updated bases documentation that will assist them in assessing the adequacy of the licensee submittals, from the viewpoint of both the planned actions, including occupational radiation exposure, and the probable costs. The purpose of this reevaluation study is to update the needed bases documentation. This report presents the results of a review and reevaluation of the PNL 1980 decommissioning study of the Washington Public Power Supply System`s Washington Nuclear Plant Two (WNP-2), which is a boiling water reactor (BWR), located at Richland, Washington, including all identifiable factors and cost assumptions which contribute significantly to the total cost of decommissioning the plant for the DECON, SAFSTOR, and ENTOMB decommissioning alternatives. These alternatives now include an initial 5-7 year period during which time the spent fuel is stored in the spent fuel pool prior to beginning major disassembly or extended safe storage of the plant. Included for information (but not part of the license termination cost) is an estimate of the cost to demolish the decontaminated and clean structures on the site and to restore the site to a {open_quotes}green field{close_quotes} condition. This report also includes consideration of the NRC requirement that decontamination and decommissioning activities leading to termination of the nuclear license be completed within 60 years of final reactor shutdown, consideration of packaging and disposal requirements for materials whose radionuclide concentrations exceed the limits for Class C low- level waste (i.e., Greater-Than-Class C), and reflects 1993 costs for labor, materials, transport, and disposal activities. Sensitivity of the total license termination cost to the disposal costs at different low-level radioactive waste disposal sites, to different depths of contaminated concrete surface removal within the facilities, and to different transport distances is also examined.

  1. Leukemia in the proximity of a German boiling-water nuclear reactor: evidence of population exposure by chromosome studies and environmental radioactivity.

    PubMed Central

    Schmitz-Feuerhake, I; Dannheim, B; Heimers, A; Oberheitmann, B; Schröder, H; Ziggel, H

    1997-01-01

    Exceptional elevation of children's leukemia appearing 5 years after the 1983 startup of the Krümmel nuclear power plant, accompanied by a significant increase of adult leukemia cases, led to investigations of radiation exposures of the population living near the plant. The rate of dicentric chromosomes in peripheral lymphocytes of seven parents of children with leukemia and in 14 other inhabitants near the plant was significantly elevated and indicated ongoing exposures over the years of its operation. These findings led to the hypothesis that chronic reactor leakages had occurred. This assumption is support by identification of artificial radioactivity in air, rainwater, soil and vegetation by the environmental monitoring program at the nuclear power plant. Calculations of the corresponding source terms show that emissions must have been well above authorized annual limits. Bone marrow doses supposedly result primarily through incorporation of bone-seeking beta- and alpha-emitters. PMID:9467072

  2. Identification and Ranking of Phenomena Leading to Peak Cladding Temperatures in Boiling Water Reactors During Large Break Loss of Coolant Accident Transients

    SciTech Connect

    Ratnayake, Ruwan K.; Ergun, S.; Hochreiter, L.E.; Baratta, A.J.

    2002-07-01

    In the licensing and validation process of best estimate codes for the analysis of nuclear reactors and postulated accident scenarios, the identification and quantification of the calculational uncertainty is required. One of the most important aspects in this process is the identification and recognition of the crucial contributing phenomena to the overall code uncertainty. The establishment of Phenomena Identification and Ranking Tables (PIRT) provides a vehicle to assist in assessing the capabilities of the computer code, and to guide the uncertainty analysis of the calculated results. The process used in this work to identify the phenomena was reviewing both licensing and best estimate calculations, as well as experiments, which had been performed for BWR LOCA analyses. The initial PIRT was developed by a group of analysts and was compared to existing BWR LOCA PIRTs as well as BWR LOCA analyses. The initial PIRT was then independently reviewed by a second panel of experts for the selected ranking of phenomena, identification of phenomena which were ignored, as well as the basis and rationale for the ranking of the phenomena. The differences between the two groups were then resolved. PIRTs have been developed for BWR types 4 and 5/6 for the Large Break Loss of Coolant Accidents (LB-LOCA). The ranking and the corresponding rationale for each phenomenon is included in tables together with the assessed uncertainty of the code capability to predict the phenomena. (authors)

  3. Boiling water reactor uranium utilization improvement potential

    SciTech Connect

    Wei, P.; Crowther, R.L.; Fennern, L.E.; Savoia, P.J.; Specker, S.R.; Tilley, R.M.; Townsend, D.B.; Wolters, R.A.

    1980-06-01

    This report documents the results of design and operational simulation studies to assess the potential for reduction of BWR uranium requirements. The impact of the improvements on separative work requirements and other fuel cycle requirements also were evaluated. The emphasis was on analysis of the improvement potential for once-through cycles, although plutonium recycle also was evaluated. The improvement potential was analyzed for several design alternatives including axial and radial natural uranium blankets, low-leakage refueling patterns, initial core enrichment distribution optimization, reinsert of initial core discharge fuel, preplanned end-of-cycle power coastdown and feedwater temperature reduction, increased discharge burnup, high enrichment discharge fuel rod reassembly and reinsert, lattice and fuel bundle design optimization, coolant density spectral shift with flow control, reduced burnable absorber residual, boric acid for cold shutdown, six-month subcycle refueling, and applications of a once-through thorium cycle design and plutonium recycle.

  4. 21 CFR 872.6710 - Boiling water sterilizer.

    Code of Federal Regulations, 2014 CFR

    2014-04-01

    ... 21 Food and Drugs 8 2014-04-01 2014-04-01 false Boiling water sterilizer. 872.6710 Section 872.6710 Food and Drugs FOOD AND DRUG ADMINISTRATION, DEPARTMENT OF HEALTH AND HUMAN SERVICES (CONTINUED) MEDICAL DEVICES DENTAL DEVICES Miscellaneous Devices § 872.6710 Boiling water sterilizer. (a) Identification. A boiling water sterilizer is an...

  5. 21 CFR 872.6710 - Boiling water sterilizer.

    Code of Federal Regulations, 2011 CFR

    2011-04-01

    ... 21 Food and Drugs 8 2011-04-01 2011-04-01 false Boiling water sterilizer. 872.6710 Section 872...) MEDICAL DEVICES DENTAL DEVICES Miscellaneous Devices § 872.6710 Boiling water sterilizer. (a) Identification. A boiling water sterilizer is an AC-powered device that consists of a container for boiling...

  6. 21 CFR 872.6710 - Boiling water sterilizer.

    Code of Federal Regulations, 2010 CFR

    2010-04-01

    ... 21 Food and Drugs 8 2010-04-01 2010-04-01 false Boiling water sterilizer. 872.6710 Section 872...) MEDICAL DEVICES DENTAL DEVICES Miscellaneous Devices § 872.6710 Boiling water sterilizer. (a) Identification. A boiling water sterilizer is an AC-powered device that consists of a container for boiling...

  7. 21 CFR 872.6710 - Boiling water sterilizer.

    Code of Federal Regulations, 2013 CFR

    2013-04-01

    ... 21 Food and Drugs 8 2013-04-01 2013-04-01 false Boiling water sterilizer. 872.6710 Section 872...) MEDICAL DEVICES DENTAL DEVICES Miscellaneous Devices § 872.6710 Boiling water sterilizer. (a) Identification. A boiling water sterilizer is an AC-powered device that consists of a container for boiling...

  8. 21 CFR 872.6710 - Boiling water sterilizer.

    Code of Federal Regulations, 2012 CFR

    2012-04-01

    ... 21 Food and Drugs 8 2012-04-01 2012-04-01 false Boiling water sterilizer. 872.6710 Section 872...) MEDICAL DEVICES DENTAL DEVICES Miscellaneous Devices § 872.6710 Boiling water sterilizer. (a) Identification. A boiling water sterilizer is an AC-powered device that consists of a container for boiling...

  9. 77 FR 76089 - Advisory Committee on Reactor Safeguards (ACRS); Meeting of the ACRS Subcommittee on Advanced...

    Federal Register 2010, 2011, 2012, 2013, 2014

    2012-12-26

    ... 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 (ABWR... Committee on Reactor Safeguards. BILLING CODE 7590-01-P...

  10. Ground Water Rule - Boil Water Advisory - Public Notification Template

    EPA Pesticide Factsheets

    The Ground Water Rule - Boil Water Advisory - Public Notification Template can be use to issue a Tier 1 Public Notification when it has been determined that source ground water is contaminated with E. Coli bacteria.

  11. 76 FR 18585 - Advisory Committee on Reactor Safeguards (ACRS); Meeting of the ACRS Subcommittee on Advanced...

    Federal Register 2010, 2011, 2012, 2013, 2014

    2011-04-04

    ... 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 (ABWR); Notice of Meeting The ACRS Subcommittee on Advanced Boiling Water...

  12. Revised analyses of decommissioning for the reference boiling water reactor power station. Effects of current regulatory and other considerations on the financial assurance requirements of the decommissioning rule and on estimates of occupational radiation exposure - appendices. Final report

    SciTech Connect

    Smith, R.I.; Bierschbach, M.C.; Konzek, G.J.; McDuffie, P.N.

    1996-07-01

    The NRC staff is in need of decommissioning bases documentation that will assist them in assessing the adequacy of the licensee submittals, from the viewpoint of both the planned actions, including occupational radiation exposure, and the probable costs. The purpose of this reevaluation study is to update the needed bases documentation. This report presents the results of a review and reevaluation of the PNL 1980 decommissioning study of the Washington Public Power Supply System`s Washington Nuclear Plant Two (WNP-2) located at Richland, Washington, including all identifiable factors and cost assumptions which contribute significantly to the total cost of decommissioning the plant for the DECON, SAFSTOR, and ENTOMB decommissioning alternatives. These alternatives now include an initial 5-7 year period during which time the spent fuel is stored in the spent fuel pool prior to beginning major disassembly or extended safe storage of the plant. Included for information (but not presently part of the license termination cost) is an estimate of the cost to demolish the decontaminated and clear structures on the site and to restore the site to a {open_quotes}green field{close_quotes} condition. This report also includes consideration of the NRC requirement that decontamination and decommissioning activities leading to termination of the nuclear license be completed within 60 years of final reactor shutdown, consideration of packaging and disposal requirements for materials whose radionuclide concentrations exceed the limits for Class C low-level waste (i.e., Greater-Than-Class C), and reflects 1993 costs for labor, materials, transport, and disposal activities. Sensitivity of the total license termination cost to the disposal costs at different low-level radioactive waste disposal sites, to different depths of contaminated concrete surface removal within the facilities, and to different transport distances is also examined.

  13. 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 (NPP) increases, safety inspection has increased in importance. However, precisely what is important, and what is not important. What should one focus inspection efforts on. Over the last two years Probabilistic Risk Assessment (PR) 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 about 25% of the US nuclear power plants have completed PRAs, and also, inspectors want PRA-based tools for these plants. This paper describes the BNL program to develop generic BWR PRA-based inspection insights or inspection guidance designed to be applied to plants without PRAs.

  14. ABWR (advanced boiling water reactor) Design Verification Program

    SciTech Connect

    Fox, J.N.

    1990-10-01

    The ABWR Design Verification Program is aimed at restoring confidence in the US licensing process by demonstrating its workability by obtaining USNRC preapproval of GE's ABWR Standard Plant. The purpose of this work is to achieve full NRC approval of the ABWR through the award of an NRC Staff final design approval (FDA) and design certification. The approach is to (1) establish a licensing basis with the NRC Staff for the ABWR, (2) prepare and submit, for NRC Staff review, an SSAR to obtain an FDA, and (3) participate in a rulemaking process to obtain certification of the ABWR design. This program was initiated August 27, 1986. This report, the fourth annual progress report, summarizes progress on this program from October 1, 1989 through September 30, 1990. 9 refs., 5 tabs.

  15. Test Plan for the Boiling Water Reactor Dry Cask Simulator

    SciTech Connect

    Durbin, Samuel; Lindgren, Eric R.

    2015-11-01

    The thermal performance of commercial nuclear spent fuel dry storage casks are evaluated through detailed numerical analysis . These modeling efforts are completed by the vendor to demonstrate performance and regulatory compliance. The calculations are then independently verified by the Nuclear Regulatory Commission (NRC). Carefully measured data sets generated from testing of full sized casks or smaller cask analogs are widely recognized as vital for validating these models. Recent advances in dry storage cask designs have significantly increased the maximum thermal load allowed in a cask in part by increasing the efficiency of internal conduction pathways and by increasing the internal convection through greater canister helium pressure. These same vertical, canistered cask systems rely on ventilation between the canister and the overpack to convect heat away from the canister to the environment for both above and below-ground configurations. While several testing programs have been previously conducted, these earlier validation attempts did not capture the effects of elevated helium pressures or accurately portray the external convection of above-ground and below-ground canistered dry cask systems. The purpose of the investigation described in this report is to produce a data set that can be used to test the validity of the assumptions associated with the calculations presently used to determine steady-state cladding temperatures in modern vertical, canistered dry cask systems. The BWR cask simulator (BCS) has been designed in detail for both the above-ground and below-ground venting configurations. The pressure vessel representing the canister has been designed, fabricated, and pressure tested for a maximum allowable pressure (MAWP) rating of 24 bar at 400 deg C. An existing electrically heated but otherwise prototypic BWR Incoloy-clad test assembly is being deployed inside of a representative storage basket and cylindrical pressure vessel that represents the canister. The symmetric single assembly geometry with well-controlled boundary conditions simplifies interpretation of results. Various configurations of outer concentric ducting will be used to mimic conditions for above and below-ground storage configurations of vertical, dry cask systems with canisters. Radial and axial temperature profiles will be measured for a wide range of decay power and helium cask pressures. Of particular interest is the evaluation of the effect of increased helium pressure on allowable heat load and the effect of simulated wind on a simplified below ground vent configuration. While incorporating the best available information, this test plan is subject to changes due to improved understanding from modeling or from as-built deviations to designs. As-built conditions and actual procedures will be documented in the final test report.

  16. Pressure suppression containment system for boiling water reactor

    DOEpatents

    Gluntz, Douglas M.; Nesbitt, Loyd B.

    1997-01-01

    A system for suppressing the pressure inside the containment of a BWR following a postulated accident. A piping subsystem is provided which features a main process pipe that communicates the wetwell airspace to a connection point downstream of the guard charcoal bed in an offgas system and upstream of the main bank of delay charcoal beds which give extensive holdup to offgases. The main process pipe is fitted with both inboard and outboard containment isolation valves. Also incorporated in the main process pipe is a low-differential-pressure rupture disk which prevents any gas outflow in this piping whatsoever until or unless rupture occurs by virtue of pressure inside this main process pipe on the wetwell airspace side of the disk exceeding the design opening (rupture) pressure differential. The charcoal holds up the radioactive species in the noncondensable gas from the wetwell plenum by adsorption, allowing time for radioactive decay before the gas is vented to the environs.

  17. Pressure suppression containment system for boiling water reactor

    DOEpatents

    Gluntz, D.M.; Nesbitt, L.B.

    1997-01-21

    A system is disclosed for suppressing the pressure inside the containment of a BWR following a postulated accident. A piping subsystem is provided which features a main process pipe that communicates the wetwell airspace to a connection point downstream of the guard charcoal bed in an offgas system and upstream of the main bank of delay charcoal beds which give extensive holdup to offgases. The main process pipe is fitted with both inboard and outboard containment isolation valves. Also incorporated in the main process pipe is a low-differential-pressure rupture disk which prevents any gas outflow in this piping whatsoever until or unless rupture occurs by virtue of pressure inside this main process pipe on the wetwell airspace side of the disk exceeding the design opening (rupture) pressure differential. The charcoal holds up the radioactive species in the noncondensable gas from the wetwell plenum by adsorption, allowing time for radioactive decay before the gas is vented to the environs. 3 figs.

  18. Passive Gamma Analysis of the Boiling-Water-Reactor Assemblies

    SciTech Connect

    Vo, Duc Ta; Favalli, Andrea

    2016-03-31

    Passive gamma analysis can be used to determine BU and CT of BWR assembly. The analysis is somewhat more complicated and less effective than similar method for PWR assemblies. From the measurements along the lengths of the BWR1 and BWR9 assemblies, there are hints that we may be able to use their information to help improve the model functions for better results.

  19. NUCLEAR REACTOR

    DOEpatents

    Treshow, M.

    1961-09-01

    A boiling-water nuclear reactor is described wherein control is effected by varying the moderator-to-fuel ratio in the reactor core. This is accomplished by providing control tubes containing a liquid control moderator in the reactor core and providing means for varying the amount of control moderatcr within the control tubes.

  20. 78 FR 37595 - Advisory Committee on Reactor Safeguards (ACRS); Meeting of the ACRS Subcommittee on Advanced...

    Federal Register 2010, 2011, 2012, 2013, 2014

    2013-06-21

    ... 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 (ABWR.... Kathy Weaver, Acting Chief, Technical Support Branch, Advisory Committee on Reactor Safeguards....

  1. 76 FR 5218 - Advisory Committee on Reactor Safeguards (ACRS); Meeting of the ACRS Subcommittee on Advanced...

    Federal Register 2010, 2011, 2012, 2013, 2014

    2011-01-28

    ... COMMISSION Advisory Committee on Reactor Safeguards (ACRS); Meeting of the ACRS Subcommittee on Advanced Boiling Water Reactor (ABWR); Notice of Meeting The ACRS Subcommittee on Advanced Boiling Water Reactor... inconvenience. Dated: January 24, 2011. Antonio Dias, Chief, Reactor Safety Branch B, Advisory Committee...

  2. 77 FR 59678 - Advisory Committee on Reactor Safeguards (ACRS), Meeting of the ACRS Subcommittee on Advanced...

    Federal Register 2010, 2011, 2012, 2013, 2014

    2012-09-28

    ... 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 (ABWR... Branch, Advisory Committee on Reactor Safeguards. BILLING CODE 7590-01-P...

  3. 78 FR 20959 - Advisory Committee on Reactor Safeguards (ACRS); Meeting of the ACRS Subcommittee on Advanced...

    Federal Register 2010, 2011, 2012, 2013, 2014

    2013-04-08

    ... COMMISSION Advisory Committee on Reactor Safeguards (ACRS); Meeting of the ACRS Subcommittee on Advanced Boiling Water Reactor The ACRS Subcommittee on Advanced Boiling Water Reactor (ABWR) will hold a meeting.... Antonio Dias, Technical Advisor, Advisory Committee on Reactor Safeguards. BILLING CODE 7590-01-P...

  4. Summary and bibliography of safety-related events at boiling-water nuclear power plants as reported in 1980

    SciTech Connect

    McCormack, K.E.; Gallaher, R.B.

    1982-03-01

    This document presents a bibliography that contains 100-word abstracts of event reports submitted to the US Nuclear Regulatory Commission concerning operational events that occurred at boiling-water-reactor nuclear power plants in 1980. The 1547 abstracts included on microfiche in this bibliography describe incidents, failures, and design or construction deficiencies that were experienced at the facilities. These abstracts are arranged alphabetically by reactor name and then chronologically for each reactor. Full-size keyword and permuted-title indexes to facilitate location of individual abstracts are provided following the text. Tables that summarize the information contained in the bibliography are also provided. The information in the tables includes a listing of the equipment items involved in the reported events and the associated number of reports for each item. Similar information is given for the various kinds of instrumentation and systems, causes of failures, deficiencies noted, and the time of occurrence (i.e., during refueling, operation, testing, or construction).

  5. Children's Understanding of Changes of State Involving the Gas State, Part 1: Boiling Water and the Particle Theory.

    ERIC Educational Resources Information Center

    Johnson, Philip

    1998-01-01

    Explores the development of children's conception of a substance and reports the findings in relation to children's understanding of boiling water and particle ideas. Argues that boiling water should have a broad significance in the curriculum. Contains 23 references. (DDR)

  6. 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.

  7. 76 FR 34276 - Advisory Committee on Reactor Safeguards (ACRS), Meeting of the ACRS Subcommittee on Advanced...

    Federal Register 2010, 2011, 2012, 2013, 2014

    2011-06-13

    ... 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 (ABWR) will hold a meeting on June 21, 2011, Room T-2B1,...

  8. 76 FR 68793 - Advisory Committee on Reactor Safeguards (ACRS) Meeting of the ACRS Subcommittee on Economic...

    Federal Register 2010, 2011, 2012, 2013, 2014

    2011-11-07

    ... No: 2011-28737] NUCLEAR REGULATORY COMMISSION Advisory Committee on Reactor Safeguards (ACRS) Meeting of the ACRS Subcommittee on Economic Simplified Boiling Water Reactor; Notice of Meeting The ACRS Subcommittee on Economic Simplified Boiling Water Reactor (ESBWR) will hold a meeting on November 30,...

  9. 76 FR 27102 - Advisory Committee on Reactor Safeguards (ACRS) Meeting of the ACRS Subcommittee on Economic...

    Federal Register 2010, 2011, 2012, 2013, 2014

    2011-05-10

    ... COMMISSION Advisory Committee on Reactor Safeguards (ACRS) Meeting of the ACRS Subcommittee on Economic Simplified Boiling Water Reactor; Notice of Meeting The ACRS Subcommittee on Economic Simplified Boiling Water Reactor (ESBWR) will hold a meeting on May 26, 2011, Room T-2B1, 11545 Rockville Pike,...

  10. 76 FR 62866 - Advisory Committee on Reactor Safeguards (ACRS) Meeting of the ACRS Subcommittee on Economic...

    Federal Register 2010, 2011, 2012, 2013, 2014

    2011-10-11

    ... COMMISSION Advisory Committee on Reactor Safeguards (ACRS) Meeting of the ACRS Subcommittee on Economic Simplified Boiling Water Reactor; Notice of Meeting The ACRS Subcommittee on Economic Simplified Boiling Water Reactor (ESBWR) will hold a meeting on October 21, 2011, Room T-2B1, 11545 Rockville...

  11. Issue a Boil-Water Advisory or Wait for Definitive Information? A Decision Analysis

    PubMed Central

    Wagner, Michael M.; Wallstrom, Garrick L.; Onisko, Agnieszka

    2005-01-01

    Objective Study the decision to issue a boil-water advisory in response to a spike in sales of diarrhea remedies or wait 72 hours for the results of definitive testing of water and people. Methods Decision analysis. Results In the base-case analysis, the optimal decision is test-and-wait. If the cost of issuing a boil-water advisory is less than 13.92 cents per person per day, the optimal decision is to issue the boil-water advisory immediately. Conclusions Decisions based on surveillance data that are suggestive but not conclusive about the existence of a disease outbreak can be modeled. PMID:16779145

  12. 75 FR 26967 - Guidance for Industry: Use of Water by Food Manufacturers in Areas Subject to a Boil-Water...

    Federal Register 2010, 2011, 2012, 2013, 2014

    2010-05-13

    ... HUMAN SERVICES Food and Drug Administration Guidance for Industry: Use of Water by Food Manufacturers in Areas Subject to a Boil-Water Advisory; Availability AGENCY: Food and Drug Administration, HHS. ACTION... entitled ``Guidance for Industry: Use of Water by Food Manufacturers in Areas Subject to a Boil-Water...

  13. 75 FR 55365 - Advisory Committee on Reactor Safeguards (ACRS) Meeting of the ACRS Joint Subcommittee

    Federal Register 2010, 2011, 2012, 2013, 2014

    2010-09-10

    ... From the Federal Register Online via the Government Publishing Office NUCLEAR REGULATORY COMMISSION Advisory Committee on Reactor Safeguards (ACRS) Meeting of the ACRS Joint Subcommittee The ACRS Subcommittees on Thermal Hydraulics Phenomena; Advanced Boiling Water Reactor (ABWR); and Materials,...

  14. Boiling water scarification plus stratification improves germination of Iliamna rivularis (Malvaceae) seeds

    Treesearch

    Katri Himanen; Markku Nygren; R. Kasten Dumroese

    2012-01-01

    Scarification with boiling water plus stratification was most effective in improving germination of Iliamna rivularis (Douglas ex Hook.) Greene (Malvaceae) in an experiment that compared 3 treatments. Seeds from 15 sites representing 5 western US states were used in the experiment. Initial response of the seedlots to the treatments was similar, apart from one seedlot....

  15. A community waterborne outbreak of salmonellosis and the effectiveness of a boil water order.

    PubMed Central

    Angulo, F J; Tippen, S; Sharp, D J; Payne, B J; Collier, C; Hill, J E; Barrett, T J; Clark, R M; Geldreich, E E; Donnell, H D; Swerdlow, D L

    1997-01-01

    OBJECTIVES: A 1993 large water-borne outbreak of Salmonella typhimurium infections in Gideon, Mo, a city of 1100 with an unchlorinated community water supply, was investigated to determine the source of contamination and the effectiveness of an order to boil water. METHODS: A survey of household members in Gideon and the surrounding township produced information on diarrheal illness, water consumption, and compliance with the boil water order. RESULTS: More than 650 persons were ill; 15 were hospitalized, and 7 died. Persons consuming city water were more likely to be ill (relative risk [RR] = 9.1, 95% confidence interval [CI] = 2.9, 28.4), and the attack rate increased with increased water consumption. S. typhimurium was recovered from samples taken from a city fire hydrant and a water storage tower. Persons in 31% (30/ 98) of city households had drunk unboiled water after being informed about the boil water order, including 14 individuals who subsequently became ill. Reasons for noncompliance included "not remembering" (44%) and "disbelieving" (25%) the order. CONCLUSIONS: Communities with deteriorating water systems risk widespread illness unless water supplies are properly operated and maintained. Effective education to improve compliance during boil water orders is needed. PMID:9146435

  16. Revised analyses of decommissioning for the reference boiling water reactor power station. Effects of current regulatory and other considerations on the financial assurance requirements of the decommissioning rule and on estimates of occupational radiation exposure: Appendices, draft report for comment. Volume 2

    SciTech Connect

    Smith, R.I.; Bierschbach, M.C.; Konzek, G.J.

    1994-09-01

    On June 27, 1988, the U.S. Nuclear Regulatory Commission (NRC) published in the Federal Register (53 FR 24018) the final rule for the General Requirements for Decommissioning Nuclear Facilities. With the issuance of the final rule, owners and operators of licensed nuclear power plants are required to prepare, and submit to the NRC for review, decommissioning plans and cost estimates. The NRC staff is in need of updated bases documentation that will assist them in assessing the adequacy of the licensee submittals, from the viewpoint of both the planned actions, including occupational radiation exposure, and the probable costs. The purpose of this reevaluation study is to update the needed bases documentation. This report presents the results of a review and reevaluation of the PNL 1980 decommissioning study of the Washington Public Power Supply System`s WNP-2, including all identifiable factors and cost assumptions which contribute significantly to the total cost of decommissioning the plant for the DECON, SAFSTOR, and ENTOMB decommissioning alternatives, which now include an initial 5-7 year period during which time the spent fuel is stored in the spent fuel pool prior to beginning major disassembly or extended safe storage of the plant. This report also includes consideration of the NRC requirement that decontamination and decommissioning activities leading to termination of the nuclear license be completed within 60 years of final reactor shutdown, consideration of packaging and disposal requirements for materials whose radionuclide concentrations exceed the limits for Class C low-level waste. Costs for labor, materials, transport, and disposal activities are given in 1993 dollars. Sensitivities of the total license termination cost to the disposal costs at different low-level radioactive waste disposal sites, to different depths of contaminated concrete surface removal within the facilities, and to different transport distances are also examined.

  17. 78 FR 46378 - La Crosse Boiling Water Reactor, Environmental Assessment and Finding of No Significant Impact...

    Federal Register 2010, 2011, 2012, 2013, 2014

    2013-07-31

    ... and efficient. However, the EP Final Rule was only an enhancement to the NRC's regulations and was not... Agencywide Documents Access and Management System (ADAMS), which provides text and image files of NRC's...

  18. Offsite dose calculation manual guidance: Standard radiological effluent controls for boiling water reactors

    SciTech Connect

    Meinke, W.W.; Essig, T.H.

    1991-04-01

    This report contains guidance which may be voluntarily used by licensees who choose to implement the provision of Generic Letter 89-- 01, which allows Radiological Effluent Technical Specifications (RETS) to be removed from the main body of the Technical Specifications and placed in the Offsite Dose Calculation Manual (ODCM). Guidance is provided for Standard Effluent Controls definitions, Controls for effluent monitoring instrumentation, Controls for effluent releases, Controls for radiological environmental monitoring, and the basis for Controls. Guidance on the formulation of RETS has been available in draft form for a number of years; the current effort simply recasts those RETS into Standard Radiological Effluent Controls for application to the ODCM. 11 tabs.

  19. Accident source terms for boiling water reactors with high burnup cores.

    SciTech Connect

    Gauntt, Randall O.; Powers, Dana Auburn; Leonard, Mark Thomas

    2007-11-01

    The primary objective of this report is to provide the technical basis for development of recommendations for updates to the NUREG-1465 Source Term for BWRs that will extend its applicability to accidents involving high burnup (HBU) cores. However, a secondary objective is to re-examine the fundamental characteristics of the prescription for fission product release to containment described by NUREG-1465. This secondary objective is motivated by an interest to understand the extent to which research into the release and behaviors of radionuclides under accident conditions has altered best-estimate calculations of the integral response of BWRs to severe core damage sequences and the resulting radiological source terms to containment. This report, therefore, documents specific results of fission product source term analyses that will form the basis for the HBU supplement to NUREG-1465. However, commentary is also provided on observed differences between the composite results of the source term calculations performed here and those reflected NUREG-1465 itself.

  20. Technology, safety and costs of decommissioning a reference boiling water reactor power station. Appendices. Volume 2

    SciTech Connect

    Oak, H.D.; Holter, G.M.; Kennedy, W.E. Jr.; Konzek, G.J.

    1980-06-01

    Appendices are presented concerning the evaluations of decommissioning financing alternatives; reference site description; reference BWR facility description; radiation dose rate and concrete surface contamination data; radionuclide inventories; public radiation dose models and calculated maximum annual doses; decommissioning methods; generic decommissioning information; immediate dismantlement details; passive safe storage, continuing care, and deferred dismantlement details; entombment details; demolition and site restoration details; cost estimating bases; public radiological safety assessment details; and details of alternate study bases.

  1. Source term attenuation by water in the Mark I boiling water reactor drywell

    SciTech Connect

    Powers, D.A.

    1993-09-01

    Mechanistic models of aerosol decontamination by an overlying water pool during core debris/concrete interactions and spray removal of aerosols from a Mark I drywell atmosphere are developed. Eighteen uncertain features of the pool decontamination model and 19 uncertain features of the model for the rate coefficient of spray removal of aerosols are identified. Ranges for values of parameters that characterize these uncertain features of the models are established. Probability density functions for values within these ranges are assigned according to a set of rules. A Monte Carlo uncertainty analysis of the decontamination factor produced by water pools 30 and 50 cm deep and subcooled 0--70 K is performed. An uncertainty analysis for the rate constant of spray removal of aerosols is done for water fluxes of 0.25, 0.01, and 0.001 cm{sup 3} H{sub 2}O/cm{sup 2}-s and decontamination factors of 1.1, 2, 3.3, 10, 100, and 1000.

  2. 77 FR 27097 - LaCrosse Boiling Water Reactor, Exemption From Certain Requirements, Vernon County, WI

    Federal Register 2010, 2011, 2012, 2013, 2014

    2012-05-08

    ... passed. The shutdown status of LACBWR means that there are no longer interconnected operating systems... Document Access and Management System (ADAMS), which provides text and image files of NRC's public...

  3. 76 FR 3540 - U.S. Advanced Boiling Water Reactor Aircraft Impact Design Certification Amendment

    Federal Register 2010, 2011, 2012, 2013, 2014

    2011-01-20

    ...: Rulemakings and Adjudications Staff. E-mail comments to: Rulemaking.Comments@nrc.gov . If you do not receive a... accessing the documents located in ADAMS, contact the NRC's PDR reference staff at 1-800-397-4209, or 301... that the applicant has complied with 10 CFR 50.150. Specifically, the staff confirmed that...

  4. Results for the Aboveground Configuration of the Boiling Water Reactor Dry Cask Simulator

    SciTech Connect

    Durbin, Samuel G.; Lindgren, Eric R.

    2016-09-30

    The thermal performance of commercial nuclear spent fuel dry storage casks is evaluated through detailed numerical analysis. These modeling efforts are completed by the vendor to demonstrate performance and regulatory compliance. The calculations are then independently verified by the Nuclear Regulatory Commission (NRC). Carefully measured data sets generated from testing of full sized casks or smaller cask analogs are widely recognized as vital for validating these models. Recent advances in dry storage cask designs have significantly increased the maximum thermal load allowed in a cask in part by increasing the efficiency of internal conduction pathways and also by increasing the internal convection through greater canister helium pressure. These same canistered cask systems rely on ventilation between the canister and the overpack to convect heat away from the canister to the environment for both above and belowground configurations. While several testing programs have been previously conducted, these earlier validation attempts did not capture the effects of elevated helium pressures or accurately portray the external convection of aboveground and belowground canistered dry cask systems. The purpose of the current investigation was to produce data sets that can be used to test the validity of the assumptions associated with the calculations used to determine steady-state cladding temperatures in modern dry casks that utilize elevated helium pressure in the sealed canister in an aboveground configuration.

  5. 77 FR 36014 - Initial Test Program of Emergency Core Cooling Systems for Boiling-Water Reactors

    Federal Register 2010, 2011, 2012, 2013, 2014

    2012-06-15

    ... NRC-2012-0134. You may submit comments by the following methods: Federal rulemaking Web site: Go to... following methods: Federal Rulemaking Web Site: Go to http://www.regulations.gov and search for Docket ID... begin the search, select ``ADAMS Public Documents'' and then select ``Begin Web- based ADAMS...

  6. 76 FR 78096 - U.S. Advanced Boiling Water Reactor Aircraft Impact Design Certification Amendment

    Federal Register 2010, 2011, 2012, 2013, 2014

    2011-12-16

    ..., the ``departure'' concept itself may be regarded as movement away from standardization. The GEH did...'' approach embodies the standardization concept more closely than the commenter's proposed use of departures... undermines the protection afforded by the design certification rulemaking concept. The comment included...

  7. BWR (boiling water reactor) lattice analysis using true geometry as compared to CPM-2

    SciTech Connect

    Knott, D.; Baratta, A. )

    1989-11-01

    Conventional lattice physics codes perform the two-dimensional transport calculation using an approximated geometry, whereby all pin cells are homogenized following the spectral calculations. To better calculate the true flux within the gadolinia cells, the two-dimensional transport calculation is performed in the true geometry of the lattice using KRAM, a two-dimensional characteristics solution of the transport equation. Normal fuel cells are modeled using three regions (cylindrical fuel and clad regions within a square moderator region), while fuel regions containing gad are subdivided into many annular subregions, thereby better approximating the self-shielding effect of the gad. The characteristics method breaks the system being analyzed into regions of constant flux, as in collision probability methods, and it is, therefore, desirable to minimize the size of these regions. To this end, then, each pin cell is further subdivided diagonally into quadrants.

  8. Subchannel analysis applied to neutron noise investigation in a boiling water reactor

    SciTech Connect

    Khan, H.J.

    1986-01-01

    Drift flux parameters were developed for subchannel configuration of BWR rod bundle geometry in order to calculate the void fraction and void propagation velocities inside the individual subchannels. Void fraction calculations were performed using subchannel drift flux code CANAL. Using void and flow distributions, values of the distribution parameters were estimated for bundle averaged void fraction calculation. The simple geometries taken into consideration are circular tubes, rectangular channels, and parallel plates. A C/sub 0/ model was also developed for parallel subchannel configuration. Successful prediction of average void fraction is observed for the annular rod bundle geometry of the FRIGG experiment. A C/sub 0/ model developed for one-dimensional subchannel geometry was applied to annular subchannels using Zuber-Findlay's void quality model. This is the first time subchannel void fraction is predicted by subchannel drift flux parameters. It appears that the magnitude of C/sub 0/ varies between subchannels but remains almost constant within each subchannel. A two-dimensional derivation of C/sub 0/ for the three types of subchannel (corner, side, and center) of BWR rod-bundle geometry was developed.

  9. Assessment of severe accident prevention and mitigation features: BWR (boiling water reactor), Mark I containment design

    SciTech Connect

    Pratt, W.T.; Eltawila, F.; Perkins, K.R.; Fitzpatrick, R.G.; Luckas, W.J.; Lehner, J.R.; Davis, P.

    1988-07-01

    Plant features and operator actions, which have been found to be important in either preventing or mitigating severe accidents in BWRs with Mark I containments (BWR Mark I's) have been identified. These features and actions were developed from insights derived from reviews of in-depth risk assessments performed specifically for the Peach Bottom plant and from assessment of other relevant studies. Accident sequences that dominate the core-damage frequency and those accident sequences that are of potentially high consequence were identified. Vulnerabilities of the BWR Mark I to severe accident containment loads were also identified. In addition, those features of a BWR Mark I, which are important for preventing core damage and are available for mitigating fission-product release to the environment were also identified. This report is issued to provide focus to an analyst examining an individual plant. This report calls attention to plant features and operator actions and provides a list of deterministic attributes for assessing those features and actions found to be helpful in reducing the overall risk for Peach Bottom and other Mark I plants. Thus, the guidance is offered as a resource in examining the subject plant to determine if the same, or similar, plant features and operator actions will be of value in reducing overall plant risk. This report is intended to serve solely as guidance.

  10. Apparatus for draining lower drywell pool water into suppresion pool in boiling water reactor

    DOEpatents

    Gluntz, Douglas M.

    1996-01-01

    An apparatus which mitigates temperature stratification in the suppression pool water caused by hot water drained into the suppression pool from the lower drywell pool. The outlet of a spillover hole formed in the inner bounding wall of the suppression pool is connected to and in flow communication with one end of piping. The inlet end of the piping is above the water level in the suppression pool. The piping is routed down the vertical downcomer duct and through a hole formed in the thin wall separating the downcomer duct from the suppression pool water. The piping discharge end preferably has an elevation at or near the bottom of the suppression pool and has a location in the horizontal plane which is removed from the point where the piping first emerges on the suppression pool side of the inner bounding wall of the suppression pool. This enables water at the surface of the lower drywell pool to flow into and be discharged at the bottom of the suppression pool.

  11. 77 FR 38339 - Dairyland Power Cooperative, La Crosse Boiling Water Reactor Exemption From Certain Security...

    Federal Register 2010, 2011, 2012, 2013, 2014

    2012-06-27

    ... series of demonstration plants funded, in part, by the U.S. Atomic Energy Commission (AEC). The nuclear... risk to the public health and safety.'' The U.S. Nuclear Regulatory Commission (NRC or the Commission... is developing an onsite independent spent fuel storage installation (ISFSI) and plans to move...

  12. 77 FR 38338 - Dairyland Power Cooperative; La Crosse Boiling Water Reactor Exemption From Certain Security...

    Federal Register 2010, 2011, 2012, 2013, 2014

    2012-06-27

    ... series of demonstration plants funded, in part, by the U.S. Atomic Energy Commission (AEC). The nuclear... risk to the public health and safety.'' The U.S. Nuclear Regulatory Commission (NRC or the Commission... 18th day of June 2012. For the U.S. Nuclear Regulatory Commission. Keith I. McConnell, Deputy...

  13. Experimental Study on Performance of a Box Solar Cooker with Flat Plate Collector to Boil Water

    NASA Astrophysics Data System (ADS)

    Sitepu, T.; Gunawan, S.; Nasution, D. M.; Ambarita, H.; Siregar, R. E. T.; Ronowikarto, A. D.

    2017-03-01

    In this study, a flat plate type solar cooker is tested by exposing in solar irradiation. The objective is to examine the performance of solar cooker in boiling water. The solar cooker is a box type with collector area and height are 100 × 100 cm and 40 cm, respectively. Vessel for water is made of aluminum plate with diameter and height of 22 cm and 15 cm. The experiments are performed by varying mass of the water. It is 2 kg and 4 kg, respectively. Every experiment starts from 10:00 AM until the boiling temperature is reached. The parameters measured are radiance intensity, ambient and solar box cooker temperatures, and wind speed. The results show that the duration of water heating up to 100°C with water mass 2 kg within 2 hours 45 minutes and water mass 4 kg within 3 hours 17 minutes. The maximum temperatur of solar box cooker is 117°C at 12:56 PM and maximum efficiency is 46.30%. The main conclusion can be drawn here is that a simple solar box cooker can be used to boil water.

  14. 75 FR 57302 - Advisory Committee on Reactor Safeguards; Public Meeting

    Federal Register 2010, 2011, 2012, 2013, 2014

    2010-09-20

    ... COMMISSION Advisory Committee on Reactor Safeguards; Public Meeting In accordance with the purposes of Sections 29 and 182b of the Atomic Energy Act (42 U.S.C. 2039, 2232b), the Advisory Committee on Reactor... Associated with the Economic Simplified Boiling Water Reactor (ESBWR) Design Certification Application...

  15. METHOD AND APPARATUS FOR CONTROLLING DIRECT-CYCLE NEUTRONIC REACTORS

    DOEpatents

    Reed, G.A.

    1961-01-10

    A control arrangement is offered for a boiling-water reactor. Boric acid is maintained in the water in the reactor and the amount in the reactor is controlled by continuously removing a portion of the water from the reactor, concentrating the boric acid by evaporating the water therefrom, returning a controlled amount of the acid to the reactor, and simultaneously controlling the water level by varying the rate of spent steam return to the reactor.

  16. 75 FR 10840 - Advisory Committee on Reactor Safeguards (ACRS) Meeting of the Subcommittee on Advanced Boiling...

    Federal Register 2010, 2011, 2012, 2013, 2014

    2010-03-09

    ... COMMISSION Advisory Committee on Reactor Safeguards (ACRS) Meeting of the Subcommittee on Advanced Boiling Water Reactor (ABWR); Notice of Meeting The ACRS Subcommittee on ABWR will hold a meeting on March 18... 3, 2010. Antonio F. Dias, Chief, Reactor Safety Branch B, Advisory Committee on Reactor...

  17. 75 FR 7632 - Advisory Committee on Reactor Safeguards (ACRS) Meeting of the Subcommittee on Advanced Boiling...

    Federal Register 2010, 2011, 2012, 2013, 2014

    2010-02-22

    ... COMMISSION Advisory Committee on Reactor Safeguards (ACRS) Meeting of the Subcommittee on Advanced Boiling Water Reactor (ABWR) The ACRS Subcommittee on ABWR will hold a meeting on March 2, 2010, at 11545...: February 12, 2010. Antonio F. Dias, Chief Reactor Safety Branch B, Advisory Committee on Reactor...

  18. Folds and Etudes

    ERIC Educational Resources Information Center

    Bean, Robert

    2007-01-01

    In this article, the author talks about "Folds" and "Etudes" which are images derived from anonymous typing exercises that he found in a used copy of "Touch Typing Made Simple". "Etudes" refers to the musical tradition of studies for a solo instrument, which is a typewriter. Typing exercises are repetitive attempts to type words and phrases…

  19. 75 FR 21046 - Advisory Committee on Reactor Safeguards

    Federal Register 2010, 2011, 2012, 2013, 2014

    2010-04-22

    ... COMMISSION Advisory Committee on Reactor Safeguards In accordance with the purposes of Sections 29 and 182b of the Atomic Energy Act (42 U.S.C. 2039, 2232b), the Advisory Committee on Reactor Safeguards (ACRS... discussions with the NRC Chairman to discuss topics of mutual interest. 1 p.m.-4 p.m.: Boiling Water...

  20. Ultrasonic level and temperature sensor for power reactor applications

    SciTech Connect

    Dress, W.B.: Miller, G.N.

    1983-01-01

    An ultrasonic waveguide employing torsional and extensional acoustic waves has been developed for use as a level and temperature sensor in pressurized and boiling water nuclear power reactors. Features of the device include continuous measurement of level, density, and temperature producing a real-time profile of these parameters along a chosen path through the reactor vessel.

  1. 77 FR 64563 - Advisory Committee on Reactor Safeguards; Notice of Meeting

    Federal Register 2010, 2011, 2012, 2013, 2014

    2012-10-22

    ... COMMISSION Advisory Committee on Reactor Safeguards; Notice of Meeting In accordance with the purposes of Sections 29 and 182b of the Atomic Energy Act (42 U.S.C. 2039, 2232b), the Advisory Committee on Reactor... the Advanced Boiling Water Reactor (ABWR) Design for South Texas Project Units 3 and 4 (STP 3 and...

  2. 78 FR 37849 - Advisory Committee on Reactor Safeguards; Notice of Meeting

    Federal Register 2010, 2011, 2012, 2013, 2014

    2013-06-24

    ... COMMISSION Advisory Committee on Reactor Safeguards; Notice of Meeting In accordance with the purposes of Sections 29 and 182b of the Atomic Energy Act (42 U.S.C. 2039, 2232b), the Advisory Committee on Reactor... Mark I boiling water reactor. 4:15 P.M.-7:00 p.m.: Preparation of ACRS Reports (Open)--The...

  3. 77 FR 69900 - Advisory Committee on Reactor Safeguards; Notice of Meeting

    Federal Register 2010, 2011, 2012, 2013, 2014

    2012-11-21

    ... COMMISSION Advisory Committee on Reactor Safeguards; Notice of Meeting In accordance with the purposes of Sections 29 and 182b of the Atomic Energy Act (42 U.S.C. 2039, 2232b), the Advisory Committee on Reactor...) Venting Systems for Boiling Water Reactors (BWRs) with Mark I and Mark II Containment Designs, and...

  4. 78 FR 18375 - Advisory Committee on Reactor Safeguards; Notice of Meeting

    Federal Register 2010, 2011, 2012, 2013, 2014

    2013-03-26

    ... COMMISSION Advisory Committee on Reactor Safeguards; Notice of Meeting In accordance with the purposes of Sections 29 and 182b of the Atomic Energy Act (42 U.S.C. 2039, 2232b), the Advisory Committee on Reactor... the Advanced Boiling Water Reactor (ABWR) Core Design. Note: A portion of this session may be...

  5. Liquid level, void fraction, and superheated steam sensor for nuclear-reactor cores. [PWR; BWR

    DOEpatents

    Tokarz, R.D.

    1981-10-27

    This disclosure relates to an apparatus for monitoring the presence of coolant in liquid or mixed liquid and vapor, and superheated gaseous phases at one or more locations within an operating nuclear reactor core, such as pressurized water reactor or a boiling water reactor.

  6. Nuclear Power from Fission Reactors. An Introduction.

    ERIC Educational Resources Information Center

    Department of Energy, Washington, DC. Technical Information Center.

    The purpose of this booklet is to provide a basic understanding of nuclear fission energy and different fission reaction concepts. Topics discussed are: energy use and production, current uses of fuels, oil and gas consumption, alternative energy sources, fossil fuel plants, nuclear plants, boiling water and pressurized water reactors, the light…

  7. Oxidation of SnO to SnO{sub 2} thin films in boiling water at atmospheric pressure

    SciTech Connect

    Nose, K. Suzuki, A. Y.; Oda, N.; Kamiko, M.; Mitsuda, Y.

    2014-03-03

    We demonstrated that SnO is oxidized to SnO{sub 2} in boiling water. (001)-oriented SnO thin films were pulsed-laser deposited onto a glass substrate. The Sn valence number changed from (II) to (IV) by keeping SnO films in boiling water at atmospheric pressure for 5 h. Optical transparency of the obtained SnO{sub 2} films was greater than 95% in the visible light range. The SnO{sub 2} films possessed an amorphous structure, and exhibited dielectric properties. Atomic force microscopy and Fourier transform infrared spectroscopy revealed granular structures and the existence of –OH groups, which may account for the diffusion of oxidants within the film.

  8. HORIZONTAL BOILING REACTOR SYSTEM

    DOEpatents

    Treshow, M.

    1958-11-18

    Reactors of the boiling water type are described wherein water serves both as the moderator and coolant. The reactor system consists essentially of a horizontal pressure vessel divided into two compartments by a weir, a thermal neutronic reactor core having vertical coolant passages and designed to use water as a moderator-coolant posltioned in one compartment, means for removing live steam from the other compartment and means for conveying feed-water and water from the steam compartment to the reactor compartment. The system further includes auxiliary apparatus to utilize the steam for driving a turbine and returning the condensate to the feed-water inlet of the reactor. The entire system is designed so that the reactor is self-regulating and has self-limiting power and self-limiting pressure features.

  9. 77 FR 45699 - Advisory Committee On Reactor Safeguards (ACRS); Meeting of the ACRS Subcommittee On Economic...

    Federal Register 2010, 2011, 2012, 2013, 2014

    2012-08-01

    ... [Federal Register Volume 77, Number 148 (Wednesday, August 1, 2012)] [Notices] [Pages 45699-45700] [FR Doc No: 2012-18759] NUCLEAR REGULATORY COMMISSION Advisory Committee On Reactor Safeguards (ACRS); Meeting of the ACRS Subcommittee On Economic Simplified Boiling Water Reactors (ESBWR); Notice of...

  10. 75 FR 30077 - Advisory Committee on Reactor Safeguards (ACRS); Meeting of the ACRS Subcommittee on ESBWR

    Federal Register 2010, 2011, 2012, 2013, 2014

    2010-05-28

    ... 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 ESBWR The ACRS Subcommittee on Economic Simplified Boiling Water Reactor (ESBWR) will hold a meeting on June...

  11. 75 FR 32229 - Advisory Committee on Reactor Safeguards (ACRS); Meeting of the ACRS Subcommittee on ABWR

    Federal Register 2010, 2011, 2012, 2013, 2014

    2010-06-07

    ... 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 ABWR The ACRS Subcommittee on Advanced Boiling Water Reactor (ABWR) will hold a meeting on June 23-24, 2010, Room...

  12. 75 FR 57536 - Advisory Committee on Reactor Safeguards (ACRS); Meeting of the ACRS Subcommittee on ABWR

    Federal Register 2010, 2011, 2012, 2013, 2014

    2010-09-21

    ... 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 ABWR The ACRS Subcommittee on Advanced Boiling Water Reactor (ABWR) will hold a meeting on October 20, 2010, Room...

  13. 77 FR 64148 - Advisory Committee on Reactor Safeguards (ACRS) Meeting of the ACRS Subcommittee on Regulatory...

    Federal Register 2010, 2011, 2012, 2013, 2014

    2012-10-18

    ... (RG) RG 1.79, ````Preoperational Testing of Emergency Core Cooling Systems for Pressurized Water Reactors,'' Revision 2 and RG 1.79.1, ``Initial Test Program of Emergency Core Cooling Systems for Boiling-Water Reactors,'' Revision 0 (DG-1277).'' The Subcommittee will hear presentations by and...

  14. 75 FR 66803 - Advisory Committee on Reactor Safeguards (ACRS); Meeting of the ACRS Subcommittee on Advanced...

    Federal Register 2010, 2011, 2012, 2013, 2014

    2010-10-29

    ... 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 (ABWR); Notice of Meeting The ACRS Subcommittee on ABWR will hold a meeting...

  15. United States Department of Energy`s reactor core protection evaluation methodology for fires at RBMK and VVER nuclear power plants. Revision 1

    SciTech Connect

    1997-06-01

    This document provides operators of Soviet-designed RBMK (graphite moderated light water boiling water reactor) and VVER (pressurized light water reactor) nuclear power plants with a systematic Methodology to qualitatively evaluate plant response to fires and to identify remedies to protect the reactor core from fire-initiated damage.

  16. 78 FR 35990 - All Operating Boiling-Water Reactor Licensees With Mark I And Mark II Containments; Docket Nos...

    Federal Register 2010, 2011, 2012, 2013, 2014

    2013-06-14

    ... licensees' severe accident management capabilities and filtering strategies to limit the release of... licensees greater capabilities to respond to severe accidents and limit the uncontrolled release of... requirement to provide a reliable HCVS to prevent or limit core damage upon loss of heat removal capability...

  17. 10 CFR Appendix A to Part 52 - Design Certification Rule for the U.S. Advanced Boiling Water Reactor

    Code of Federal Regulations, 2012 CFR

    2012-01-01

    ... 50.2 or 52.1, or Section 11 of the Atomic Energy Act of 1954, as amended, as applicable. III. Scope... controls. E. Design activities for structures, systems, and components that are wholly outside the scope of... determined that the structures, systems, components, and design features of the U.S. ABWR design comply with...

  18. 10 CFR Appendix A to Part 52 - Design Certification Rule for the U.S. Advanced Boiling Water Reactor

    Code of Federal Regulations, 2014 CFR

    2014-01-01

    ... 50.2 or 52.1, or Section 11 of the Atomic Energy Act of 1954, as amended, as applicable. III. Scope... structures, systems, and components that are wholly outside the scope of this appendix may be performed using... structures, systems, components, and design features of the U.S. ABWR design, as contained in the GE DCD...

  19. 10 CFR Appendix A to Part 52 - Design Certification Rule for the U.S. Advanced Boiling Water Reactor

    Code of Federal Regulations, 2013 CFR

    2013-01-01

    ... 50.2 or 52.1, or Section 11 of the Atomic Energy Act of 1954, as amended, as applicable. III. Scope... structures, systems, and components that are wholly outside the scope of this appendix may be performed using... structures, systems, components, and design features of the U.S. ABWR design, as contained in the GE DCD...

  20. A New Computational Tool for Simulation of 3-D Flow and Heat Transfer in Boiling Water Reactors

    SciTech Connect

    Chen, Hudong

    2002-12-09

    This Phase I work has developed a novel hybrid Lattice Boltzmann Model for the simulation of nonideal fluid thermal dynamics and demonstrated that this model can be used to simulate fundamental two-phase flow processes including boiling initiation, bubble formation and coalescency, and flow-regime formation.

  1. FIBWR: a steady-state core flow distribution code for boiling water reactors code verification and qualification report. Final report

    SciTech Connect

    Ansari, A.F.; Gay, R.R.; Gitnick, B.J.

    1981-07-01

    A steady-state core flow distribution code (FIBWR) is described. The ability of the recommended models to predict various pressure drop components and void distribution is shown by comparison to the experimental data. Application of the FIBWR code to the Vermont Yankee Nuclear Power Station is shown by comparison to the plant measured data.

  2. 78 FR 63516 - Initial Test Program of Emergency Core Cooling Systems for New Boiling-Water Reactors

    Federal Register 2010, 2011, 2012, 2013, 2014

    2013-10-24

    ... the following methods: Federal Rulemaking Web site: Go to http://www.regulations.gov and search for... ``Begin Web- based ADAMS Search.'' For problems with ADAMS, please contact the NRC's Public Document...

  3. Safety Analysis of Small Break Loss of Coolant Accident for 1200 MWe Simplified Boiling Water Reactor (SBWR-1200 BDLB)

    SciTech Connect

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

    2002-07-01

    The objective of this research is to assess the performance of the safety systems during small break loss of coolant accident (SBLOCA) transient in the full size SBWR. RELAP5/MOD3 was used to simulate the blow-down and long-term cooling responses of the various safety systems during the accident transient. An integral test for long-term cooling under low pressure was conducted in a scaled facility with the initial conditions given by the code simulation. The code applicability and the facility scalability were evaluated by the comparison between the test data and the code simulations. The scaling analysis has been done by the comparison of the prototype code predictions and the scaled-up test data with the proper scaling multiplications and time shifting. The good agreement between the major safety parameters has shown the applicability of the RELAP5/MOD3 code and the scalability of the facility for SBWR-1200 safety analysis applications. (authors)

  4. An assessment of BWR (boiling water reactor) Mark III containment challenges, failure modes, and potential improvements in performance

    SciTech Connect

    Schroeder, J.A.; Pafford, D.J.; Kelly, D.L.; Jones, K.R.; Dallman, F.J. )

    1991-01-01

    This report describes risk-significant challenges posed to Mark III containment systems by severe accidents as identified for Grand Gulf. Design similarities and differences between the Mark III plants that are important to containment performance are summarized. The accident sequences responsible for the challenges and the postulated containment failure modes associated with each challenge are identified and described. Improvements are discussed that have the potential either to prevent or delay containment failure, or to mitigate the offsite consequences of a fission product release. For each of these potential improvements, a qualitative analysis is provided. A limited quantitative risk analysis is provided for selected potential improvements. 21 refs., 5 figs., 46 tabs.

  5. 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)

  6. 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.

  7. 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.

  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. 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)

  10. 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).

  11. NEUTRONIC REACTOR SYSTEM

    DOEpatents

    Treshow, M.

    1959-02-10

    A reactor system incorporating a reactor of the heterogeneous boiling water type is described. The reactor is comprised essentially of a core submerged adwater in the lower half of a pressure vessel and two distribution rings connected to a source of water are disposed within the pressure vessel above the reactor core, the lower distribution ring being submerged adjacent to the uppcr end of the reactor core and the other distribution ring being located adjacent to the top of the pressure vessel. A feed-water control valve, responsive to the steam demand of the load, is provided in the feedwater line to the distribution rings and regulates the amount of feed water flowing to each distribution ring, the proportion of water flowing to the submerged distribution ring being proportional to the steam demand of the load. This invention provides an automatic means exterior to the reactor to control the reactivity of the reactor over relatively long periods of time without relying upon movement of control rods or of other moving parts within the reactor structure.

  12. 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.

  13. Effect of latent heat in boiling water on the synthesis of gold nanoparticles of different sizes by using the Turkevich method.

    PubMed

    Ding, Wenchao; Zhang, Peina; Li, Yijing; Xia, Haibing; Wang, Dayang; Tao, Xutang

    2015-02-02

    The Turkevich method, involving the reduction of HAuCl4 with citrate in boiling water, allows the facile production of monodisperse, quasispherical gold nanoparticles (AuNPs). Although, it is well-known that the size of the AuNPs obtained with the same recipe vary slightly (as little as approximately 4 nm), but noticeably, from one report to another, it has rarely been studied. The present work demonstrates that this size variation can be reconciled by the small, but noticeable, effect that the latent heat in boiling water has on the size of the AuNPs obtained by using the Turkevich method. The increase in latent heat during water boiling caused an approximately 3 nm reduction in the size of the as-prepared AuNPs; this reduction in size is mainly a result of accelerated nucleation driven by the extra heat. It was further demonstrated that, the heating temperature can be utilized as an additional measure to adjust the growth rate of AuNPs during the reduction of HAuCl4 with citrate in boiling water. Therefore, the latent heat of boiling solvents may provide one way to control nucleation and growth in the synthesis of monodisperse nanoparticles.

  14. Nuclear reactor cavity decontamination machine

    SciTech Connect

    Vassalotti, M.; Obligado, A.

    1984-03-13

    Apparatus is disclosed for decontaminating the wall of a boiling water reactor cavity. A chassis on wheels is rollable on the refueling floor along the cavity curb. A pair of horizontal wheels roll against the curb. A support member extends upwardly and laterally from the chassis to clear the personnel handrail. A mast depends from the support member into the cavity and includes a horizontal reaction wheel bearing against the cavity wall. A vertically positionable carriage is mounted on the mast and carries water spray nozzles directed against the wall.

  15. Report on Reactor Physics Assessment of Candidate Accident Tolerant Fuel Cladding Materials in LWRs

    SciTech Connect

    Powers, Jeffrey J.; George, Nathan; Maldonado, G. Ivan; Worrall, Andrew

    2015-08-28

    This work focuses on ATF concepts being researched at Oak Ridge National Laboratory (ORNL), expanding on previous studies of using alternate cladding materials in pressurized water reactors (PWRs). The neutronic performance of two leading alternate cladding materials were assessed in boiling water reactors (BWRs): iron-chromium-aluminum (FeCrAl) cladding, and silicon carbide (SiC)-based composite cladding. This report fulfills ORNL Milestone M3FT-15OR0202332 within the fiscal year 2015 (FY15)

  16. Does the public receive and adhere to boil water advisory recommendations? A cross-sectional study in Newfoundland and Labrador, Canada.

    PubMed

    Jones-Bitton, Andria; Gustafson, Diana L; Butt, Kelly; Majowicz, Shannon E

    2016-01-05

    Highly publicized water supply problems highlight the importance of safe drinking water to the public. Boil water advisories (BWAs) are an important precautionary measure meant to protect public health by ensuring drinking water safety. Newfoundland and Labrador, Canada is a prime location for exploring public notification practices and adherence to recommendations as there were a total of 215 BWAs, affecting 6 % of the provincial population, in 145 communities between April 2006 and March 2007 when data for the present study were collected. Residents who received household water from a public water supply were randomly selected for a telephone interview. Collected data included participants' notification of boil water advisory, satisfaction with information provided, and their adherence to recommendations. Most participants learned that a BWA had been issued or lifted in their community through radio, television, or word of mouth. BWAs were issued for a range of operational reasons. Almost all participants who had experienced a BWA reported wanting more information about the reasons a BWA had been issued. Low adherence to water use recommendations during a BWA was common. This study is first to report on public adherence to boil water advisory recommendations in Canada. The findings raise public health concerns, particularly given the high number of BWAs issued each year. Further studies in partnership with community stakeholders and government decision-makers responsible for overseeing public water systems are needed to assess the perceptions of BWAs, the reasons for non-adherence, and to identify information dissemination methods to increase information uptake and public adherence with acceptable uses of public drinking water during a BWA.

  17. Nuclear reactor with low-level core coolant intake

    DOEpatents

    Challberg, Roy C.; Townsend, Harold E.

    1993-01-01

    A natural-circulation boiling-water reactor has skirts extending downward from control rod guide tubes to about 10 centimeters from the reactor vessel bottom. The skirts define annular channels about control rod drive housings that extend through the reactor vessel bottom. Recirculating water is forced in through the low-level entrances to these channels, sweeping bottom water into the channels in the process. The sweeping action prevents cooler water from accumulating at the bottom. This in turn minimizes thermal shock to bottom-dwelling components as would occur when accumulated cool water is swept away and suddenly replaced by warmer water.

  18. Core damage frequency (reactor design) perspectives based on IPE results

    SciTech Connect

    Camp, A.L.; Dingman, S.E.; Forester, J.A.

    1996-12-31

    This paper provides perspectives gained from reviewing 75 Individual Plant Examination (IPE) submittals covering 108 nuclear power plant units. Variability both within and among reactor types is examined to provide perspectives regarding plant-specific design and operational features, and C, modeling assumptions that play a significant role in the estimates of core damage frequencies in the IPEs. Human actions found to be important in boiling water reactors (BWRs) and in pressurized water reactors (PWRs) are presented and the events most frequently found important are discussed.

  19. Thawing dog spermatozoa in just-boiled water: submersion time and effect on sperm quality compared to thawing in water at 70 degrees C.

    PubMed

    Nöthling, J O; Dolieslager, S M J; Fillekes, R; Colenbrander, B

    2007-09-01

    Dog spermatozoa have better quality after thawing in water at 70-75 degrees C instead of 35-38 degrees C. The aim of Experiment 1 was to determine the time needed to thaw 0.5 mL straws in just-boiled (98 degrees C) water and that of Experiment 2 to determine whether thawing frozen dog spermatozoa in just-boiled water will result in better quality than thawing in water at 70 degrees C. Prior to freezing the straws of Experiment 1, a Type J thermocouple with wire diameters of 0.08 mm (Osiris Technical Systems, Centurion, South Africa) was placed in the center of each of ninety-three 0.5 mL straws (IMV Technologies, L'Aigle, France) filled with extender (Biladyl* with 0.5%, v/v of Equex STM paste**) and 54 filled with extender plus 200 x 10(6)spermatozoa/mL (Minitüb, Germany (*) and Nova Chemical Sales, MA (**)). Thirty straws with extender were thawed in water at 70 degrees C and the others in just-boiled water. Temperatures inside straws were recorded 10 times/s during warming. Two ejaculates were then collected from each of eight dogs and one from each of three others. Extended ejaculates from the same dog were pooled, frozen 8 cm above liquid nitrogen, and 2 straws from each of the 11 batches thawed in water at 70 degrees C for 8s and 2 in just-boiled water for 6.5s. Sperm morphology and viability were assessed on eosin-nigrosin smears made after thawing and the percentage progressively motile spermatozoa was estimated immediately, 1, 2 and 3h after thawing. The optimal submersion time in just-boiled water was 6.5s for both sperm concentrations, resulting in average temperatures of 23.6+/-1.5 degrees C (+/-S.E.M.) and 24.9+/-1.6 degrees C inside straws with extender or extender plus spermatozoa (P=0.6). The temperature inside straws thawed in water at 70 degrees C was 13.6+/-1.7 degrees C after 8s. Apart from a 1.5% higher (P<0.05) mean percentage motile sperm 2h after thawing, thawing dog spermatozoa in just-boiled (98 degrees C) water holds no benefit over

  20. Surface conditions of Nitinol wires, tubing, and as-cast alloys. The effect of chemical etching, aging in boiling water, and heat treatment.

    PubMed

    Shabalovskaya, S A; Anderegg, J; Laab, F; Thiel, P A; Rondelli, G

    2003-04-15

    The surface conditions of Nitinol wires and tubing were evaluated with the use of X-ray photoelectron spectroscopy, high-resolution Auger spectroscopy, electron backscattering, and scanning-electron microscopy. Samples were studied in the as-received state as well as after chemical etching, aging in boiling water, and heat treatment, and compared to a mechanically polished 600-grit-finish Nitinol surface treated similarly. General regularities in surface behavior induced by the examined surface treatments are similar for wires, tubing, and studied as-cast alloy, though certain differences in surface Ni concentration were observed. Nitinol wires and tubing from various suppliers demonstrated great variability in Ni surface concentration (0.5-15 at.%) and Ti/Ni ratio (0.4-35). The wires in the as-received state, with the exception of those with a black oxide originating in the processing procedure, revealed nickel and titanium on the surface in both elemental and oxidized states, indicating a nonpassive surface. Shape-setting heat treatment at 500 degrees C for 15 min resulted in tremendous increase in the surface Ni concentration and complete Ni oxidation. Preliminary chemical etching and boiling in water successfully prevented surface enrichment in Ni, initially resulting from heat treatment. A stoichiometric uniformly amorphous TiO(2) oxide generated during chemical etching and aging in boiling water was reconstructed at 700 degrees C, revealing rutile structure. Copyright 2003 Wiley Periodicals, Inc.

  1. Paramagnetism and improved upconversion luminescence properties of NaYF4:Yb,Er/NaGdF4 nanocomposites synthesized by a boiling water seed-mediated route

    NASA Astrophysics Data System (ADS)

    Yang, Chao-Qing; Li, Ao-Ju; Guo, Wei; Tian, Peng-Hua; Yu, Xiao-Long; Liu, Zhong-Xin; Cao, Yang; Sun, Zhong-Liang

    2016-03-01

    In a route boiling water served as reaction medium, a stoichiometric amount of rare-earth compound and fluoride are put into this system to form α-NaYF4:Yb, Er nuclei. Then prepared sample is heated at elevated temperature to improve the fluorescence intensity, and next a NaGdF4 shell grows on the surface of NaYF4 nuclei. NaYF4:Yb,Er/NaGdF4 core-shell structured upconversion nanoparticles (CSUCNPs) have been successfully synthesized by above route. The use of boiling water decreases the cubic-to-hexagonal phase transition temperature of NaYF4:Yb,Er to 350°C and increases its upconversion (UC) luminescence intensity. A heterogeneous NaGdF4 epitaxially growing on the surface of Ln3+-doped NaYF4 not only improves UC luminescence, but also creates a paramagnetic shell, which can be used as contrast agents in magnetic resonance imaging (MRI). The solution of CSUCNPs shows bright green UC fluorescence under the excitation at 980 nm in a power density only about 50 mW·cm-2. A broad spectrum with a dominant resonance at g of about 2 is observed by the electron paramagnetic resonance (EPR) spectrum of CSUCNPs. Above properties suggest that the obtained CSUCNPs could be potential candidates for dual-mode optical/magnetic bioapplications.

  2. Communication, perception and behaviour during a natural disaster involving a 'Do Not Drink' and a subsequent 'Boil Water' notice: a postal questionnaire study

    PubMed Central

    2010-01-01

    Background During times of public health emergencies, effective communication between the emergency response agencies and the affected public is important to ensure that people protect themselves from injury or disease. In order to investigate compliance with public health advice during natural disasters, we examined consumer behaviour during two water notices that were issued as a result of serious flooding. During the summer of 2007, 140,000 homes in Gloucestershire, United Kingdom, that are supplied water from Mythe treatment works, lost their drinking water for up to 17 days. Consumers were issued a 'Do Not Drink' notice when the water was restored, which was subsequently replaced with a 'Boil Water' notice. The rare occurrence of two water notices provided a unique opportunity to compare compliance with public health advice. Information source use and other factors that may affect consumer perception and behaviour were also explored. Method A postal questionnaire was sent to 1,000 randomly selected households. Chi-square, ANOVA, MANOVA and generalised estimating equation (with and without prior factor analysis) were used for quantitative analysis. Results In terms of information sources, we found high use of and clear preference for the local radio throughout the incident, but family/friends/neighbours also proved crucial at the onset. Local newspapers and the water company were associated with clarity of advice and feeling informed, respectively. Older consumers and those in paid employment were particularly unlikely to read the official information leaflets. We also found a high degree of confusion regarding which notice was in place at which time, with correct recall varying between 23.2%-26.7%, and a great number of consumers believed two notices were in place simultaneously. In terms of behaviour, overall non-compliance levels were significantly higher for the 'Do Not Drink' notice (62.9%) compared to the 'Boil Water' notice (48.3%); consumers in paid

  3. System Study: Reactor Core Isolation Cooling 1998-2014

    SciTech Connect

    Schroeder, John Alton

    2015-12-01

    This report presents an unreliability evaluation of the reactor core isolation cooling (RCIC) system at 31 U.S. commercial boiling water reactors. Demand, run hours, and failure data from fiscal year 1998 through 2014 for selected components were obtained from the Institute of Nuclear Power Operations (INPO) Consolidated Events Database (ICES). The unreliability results are trended for the most recent 10 year period, while yearly estimates for system unreliability are provided for the entire active period. No statistically significant trends were identified in the RCIC results.

  4. System Study: Reactor Core Isolation Cooling 1998–2013

    SciTech Connect

    Schroeder, John Alton

    2015-01-31

    This report presents an unreliability evaluation of the reactor core isolation cooling (RCIC) system at 31 U.S. commercial boiling water reactors. Demand, run hours, and failure data from fiscal year 1998 through 2013 for selected components were obtained from the Institute of Nuclear Power Operations (INPO) Consolidated Events Database (ICES). The unreliability results are trended for the most recent 10-year period, while yearly estimates for system unreliability are provided for the entire active period. No statistically significant trends were identified in the RCIC results.

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

    SciTech Connect

    T. E. Wierman

    2013-10-01

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

  6. Effect of chemical etching and aging in boiling water on the corrosion resistance of Nitinol wires with black oxide resulting from manufacturing process.

    PubMed

    Shabalovskaya, S; Rondelli, G; Anderegg, J; Simpson, B; Budko, S

    2003-07-15

    The effect of chemical etching in a HF/HNO(3) acid solution and aging in boiling water on the corrosion resistance of Nitinol wires with black oxide has been evaluated with the use of potentiodynamic, modified potentiostatic ASTM F746, and scratch tests. Scanning-electron microscopy, elemental XPS, and Auger analysis were employed to characterize surface alterations induced by surface treatment and corrosion testing. The effect of aging in boiling water on the temperatures of martensitic transformations and shape recovery was evaluated by means of measuring the wire electroresistance. After corrosion tests, as-received wires revealed uniformly cracked surfaces reminiscent of the stress-corrosion-cracking phenomenon. These wires exhibited negative breakdown potentials in potentiostatic tests and variable breakdown potentials in potentiodynamic tests (- 100 mV to + 400 mV versus SCE). Wires with treated surfaces did not reveal cracking or other traces of corrosion attacks in potentiodynamic tests up to + 900-1400-mV potentials and no pitting after stimulation at + 800 mV in potentiostatic tests. They exhibited corrosion behavior satisfactory for medical applications. Significant improvement of corrosion parameters was observed on the reverse scans in potentiodynamic tests after exposure of treated wires to potentials > 1000 mV. In scratch tests, the prepared surfaces repassivated only at low potentials, comparable to that of stainless steel. Tremendous improvement of the corrosion behavior of treated Nitinol wires is associated with the removal of defect surface material and the growth of stable TiO(2) oxide. The role of precipitates in the corrosion resistance of Nitinol-scratch repassivation capacity in particular-is emphasized in the discussion.

  7. Estimating boiling water reactor decommissioning costs. A user`s manual for the BWR Cost Estimating Computer Program (CECP) software: Draft report for comment

    SciTech Connect

    Bierschbach, M.C.

    1994-12-01

    With the issuance of the Decommissioning Rule (July 27, 1988), nuclear power plant licensees are required to submit to the U.S. Regulatory Commission (NRC) for review, decommissioning plans and cost estimates. This user`s manual and the accompanying Cost Estimating Computer Program (CECP) software provide a cost-calculating methodology to the NRC staff that will assist them in assessing the adequacy of the licensee submittals. The CECP, designed to be used on a personal computer, provides estimates for the cost of decommissioning BWR power stations to the point of license termination. Such cost estimates include component, piping, and equipment removal costs; packaging costs; decontamination costs; transportation costs; burial costs; and manpower costs. In addition to costs, the CECP also calculates burial volumes, person-hours, crew-hours, and exposure person-hours associated with decommissioning.

  8. Report of the US Nuclear Regulatory Commission Piping Review Committee. Volume 1. Investigation and evaluation of stress corrosion cracking in piping of boiling water reactor plants

    SciTech Connect

    Not Available

    1984-08-01

    IGSCC in BWR piping is occurring owing to a combination of material, environment, and stress factors, each of which can affect both the initiation of a stress-corrosion crack and the rate of its subsequent propagation. In evaluating long-term solutions to the problem, one needs to consider the effects of each of the proposed remedial actions. Mitigating actions to control IGSCC in BWR piping must be designed to alleviate one or more of the three synergistic factors: sensitized material, the convention BWR environment, and high tensile stresses. Because mitigating actions addressing each of these factors may not be fully effective under all anticipated operating conditions, mitigating actions should address two and preferably all three of the causative factors; e.g., material plus some control of water chemistry, or stress reversal plus controlled water chemistry.

  9. Characterization of thermal-hydraulic and ignition phenomena in prototypic, full-length boiling water reactor spent fuel pool assemblies after a complete loss-of-coolant accident.

    SciTech Connect

    Lindgren, Eric Richard; Durbin, Samuel G

    2007-04-01

    The objective of this project was to provide basic thermal-hydraulic data associated with a SFP complete loss-of-coolant accident. The accident conditions of interest for the SFP were simulated in a full-scale prototypic fashion (electrically-heated, prototypic assemblies in a prototypic SFP rack) so that the experimental results closely represent actual fuel assembly responses. A major impetus for this work was to facilitate code validation (primarily MELCOR) and reduce questions associated with interpretation of the experimental results. It was necessary to simulate a cluster of assemblies to represent a higher decay (younger) assembly surrounded by older, lower-power assemblies. Specifically, this program provided data and analysis confirming: (1) MELCOR modeling of inter-assembly radiant heat transfer, (2) flow resistance modeling and the natural convective flow induced in a fuel assembly as it heats up in air, (3) the potential for and nature of thermal transient (i.e., Zircaloy fire) propagation, and (4) mitigation strategies concerning fuel assembly management.

  10. Design, Construction and Testing of an In-Pile Loop for PWR (Pressurized Water Reactor) Simulation.

    DTIC Science & Technology

    1987-06-01

    corrosion resistance in a steam environment. For this reason zircaloy - 2 is used 109 as the primary cladding material in Boiling Water Reactors (BWR...Unfortunately, zircaloy - 2 was found to have a high affinity for monoatomic hydrogen, which formed an intermetallic compound of zirconium-hydride. The...built, the Loop duplicates the core and Steam Generator fluid surface film differential temperatures , bulk fluid temperatures , and wall fluid shear

  11. Design Study of Pb-Bi-Cooled and NaK-Cooled Small Reactors: PBWFR and DSFR

    SciTech Connect

    Otsubo, Akira; Takahashi, Minoru

    2004-07-01

    The liquid lead-bismuth eutectic (Pb-Bi) has good compatibility with water, which is different from sodium. It is expected that the Pb-Bi could be used as a coolant of the deep sea fast reactor (DSFR) and the Pb-Bi- cooled direct contact boiling water small fast reactor (PBWFR). Physics analysis of the Pb-Bi-cooled small reactor cores with and without inner control rods was performed using the computer program of General Purpose Neutronics Code System (SRAC95) developed by Japan Atomic Energy Research Institute (JAERI). The coolant of Pb-Bi seems to be good as well as NaK for small reactors. (authors)

  12. A new steam-cooled reactor

    SciTech Connect

    Schultz, M.A.; Edlund, M.C.

    1985-08-01

    A new ultrasafe type of nuclear power plant is described that has a complete ''walk-awayfrom'' characteristic. That is, the reactor can safely dissipate its shutdown heat even if its powe and water supplies are cut off. The reactor is steam cooled and is designed to operate at one fixed steam density. Its reactivity characteristics are such that if the power level increases, the steam becomes less dense than the optimum and tends to shut the reactor off. Similarly, if the reactor is flooded wit water, the reactivity greatly decreases and also shuts the reactor down. The reactor can be operated as a burner, a high-efficiency converter, or a breeder, depending on the isotopic content of the fuel. The plant operates at low pressure and relatively high efficiency with an example given at 1000 psia and 35% efficiency. The reactor is enclosed in a conventional steel vessel resembling a boiling water reactor. The vessel is connected to a large atmospheric pressure pool of water, and shutdown consists of passively cou pling the pool to the reactor through the loss of steam flow. Shutdown cooling is provided by forced air and natural draft convection cooling of the pressure vessel. Sufficient water and passive cooling are provided by the pool for many months of shutdown water cooling. The plant piping is double walled, and all paths of radiation escape, including pressure-vessel cracking, are channeled through an on-line cleanup system.

  13. (UA1 reactor fuels safety and performance)

    SciTech Connect

    Taleyarkhan, R.P.

    1990-07-13

    The traveler visited several reactor and hot cell experimental facilities connected with JAERI at the Oarai and Tokai establishments. Uranium silicide fission product release experimental data and related acquisition systems were discussed. A presentation was made by the traveler on analysis and modeling of fission product release from UAl reactor fuels. Data obtained by JAERI thus far were offered to the traveler for Oak Ridge National Laboratory (ORNL) review and analysis. This data confirmed key aspects of ORNL theoretical model predictions and will be useful for Advanced Neutron Source (ANS) design. The Oarai establishment expressed their interest and willingness to pursue ORNL/JAERI cooperative efforts in understanding volatile fission product release behavior from silicide fuels. The traveler also presented a perspective overview on ORNL severe accident analysis technology and identified areas for cooperation in JAERI's forthcoming transient testing program. JAERI staff presented plans for evaluating silicide fuel performance under transient reactivity insertion accident conditions in the Nuclear Safety Research Reactor (NSRR) facility. A surprise announcement was made concerning JAERI's most recent initiative relating to the construction of a safety demonstration reactor (SDR) at the Tokai site. The purpose of this reactor facility would be to demonstrate operational safety of both Pressurized Water Reactors (PWRs) and Boiling Water Reactors (BWRs) in support of Japan's nuclear power industry.

  14. A water contamination incident in Oslo, Norway during October 2007; a basis for discussion of boil-water notices and the potential for post-treatment contamination of drinking water supplies.

    PubMed

    Robertson, Lucy; Gjerde, Bjørn; Hansen, Elisabeth Furuseth; Stachurska-Hagen, Teresa

    2009-03-01

    Over a 5 day period in October 2007 a boil-water notice was served on the majority of Oslo, capital city of Norway, as a result of a combination of bacteriological findings (coliforms, intestinal enterococci, and E. coli), and very low numbers of Cryptosporidium oocysts and Giardia cysts in 10 L water samples taken from the water distribution network. The water source had been regularly monitored for these parasites and generally found to be negative. Over 460,000 residents were affected by the boil-water notice, as were many thousands of businesses. Despite an extensive outbreak of waterborne giardiasis in Bergen, Norway during 2004/2005, occurrence of parasites in Norwegian drinking water supplies has apparently continued to be considered to be of minimal relevance by Norwegian health authorities. Here we describe the background and occurrence of the episode in Oslo, including the species of Cryptosporidium detected, and use this event, in conjunction with incidents from other countries, as a basis to discuss the following issues: (1) under which circumstances should the occurrence of Cryptosporidium oocysts and Giardia cysts in water supplies trigger the issue of a boil-water notice, and (2) the possibilities and probabilities of post-treatment contamination events in the water distribution network.

  15. Dual-phase reactor plant with partitioned isolation condenser

    DOEpatents

    Hui, Marvin M.

    1992-01-01

    A nuclear energy plant housing a boiling-water reactor utilizes an isolation condenser in which a single chamber is partitioned into a distributor plenum and a collector plenum. Steam accumulates in the distributor plenum and is conveyed to the collector plenum through an annular manifold that includes tubes extending through a condenser pool. The tubes provide for a transfer of heat from the steam, forming a condensate. The chamber has a disk-shaped base, a cylindrical sidewall, and a semispherical top. This geometry results in a compact design that exhibits significant performance and cost advantages over prior designs.

  16. FUEL SUBASSEMBLY CONSTRUCTION FOR RADIAL FLOW IN A NUCLEAR REACTOR

    DOEpatents

    Treshow, M.

    1962-12-25

    An assembly of fuel elements for a boiling water reactor arranged for radial flow of the coolant is described. The ingress for the coolant is through a central header tube, perforated with parallel circumferertial rows of openings each having a lip to direct the coolant flow downward. Around the central tube there are a number of equally spaced concentric trays, closely fitiing the central header tube. Cylindrical fuel elements are placed in a regular pattern around the central tube, piercing the trays. A larger tube encloses the arrangement, with space provided for upward flow of coolart beyond the edge of the trays. (AEC)

  17. IGL: Internal Gap Lattice for ABWR-II Reactor

    SciTech Connect

    Kouji Hiraiwa; Yamato Hayashi; Yasushi Yamamoto; Miyuki Akiba; Takafumi Anegawa; Hiroyuki Okada

    2002-07-01

    An internal gap lattice (IGL) bundle is currently being developed under the joint study of TEPCO and Toshiba corporation for advanced boiling water reactor II (ABWR-II). IGL bundle has four sub-bundles and has a cruciform steam flow area in a center of the bundle. The IGL bundle shows favorable thermal hydraulic characteristics and reactivity performance. The reactivity performances and equilibrium core performance was evaluated based on the representative large bundle design established in ABWR-II project. By adopting the IGL design,both the reactivity in hot operation and cold shut down margin increased. In addition, absolute value of void coefficient decreased. (authors)

  18. Identification and characterization of passive safety system and inherent safety feature building blocks for advanced light-water reactors

    SciTech Connect

    Forsberg, C.W.

    1989-01-01

    Oak Ridge National Laboratory (ORNL) is investigating passive and inherent safety options for Advanced Light-Water Reactors (ALWRs). A major activity in 1989 includes identification and characterization of passive safety system and inherent safety feature building blocks, both existing and proposed, for ALWRs. Preliminary results of this work are reported herein. This activity is part of a larger effort by the US Department of Energy, reactor vendors, utilities, and others in the United States to develop improved LWRs. The Advanced Boiling Water Reactor (ABWR) program and the Advanced Pressurized Water Reactor (APWR) program have as goals improved, commercially available LWRs in the early 1990s. The Advanced Simplified Boiling Water Reactor (ASBWR) program and the AP-600 program are developing more advanced reactors with increased use of passive safety systems. It is planned that these reactors will become commercially available in the mid 1990s. The ORNL program is an exploratory research program for LWRs beyond the year 2000. Desired long-term goals for such reactors include: (1) use of only passive and inherent safety, (2) foolproof against operator errors, (3) malevolence resistance against internal sabotage and external assault and (4) walkaway safety. The acronym ''PRIME'' (Passive safety, Resilient operation, Inherent safety, Malevolence resistance, and Extended (walkaway) safety) is used to summarize these desired characteristics. Existing passive and inherent safety options are discussed in this document.

  19. 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.

  20. Russian RBMK reactor design information

    SciTech Connect

    Not Available

    1993-11-01

    This document concerns the systems, design, and operations of the graphite-moderated, boiling, water-cooled, channel-type (RBMK) reactors located in the former Soviet Union (FSU). The Russian Academy of Sciences Nuclear Safety Institute (NSI) in Moscow, Russia, researched specific technical questions that were formulated by the Pacific Northwest Laboratory (PNL) and provided detailed technical answers to those questions. The Russian response was prepared in English by NSI in a question-and-answer format. This report presents the results of that technical exchange in the context they were received from the NSI organization. Pacific Northwest Laboratory is generating this document to support the US Department of Energy (DOE) community in responding to requests from FSU states, which are seeking Western technological and financial assistance to improve the safety systems of the Russian-designed reactors. This report expands upon information that was previously available to the United States through bilateral information exchanges, international nuclear society meetings, International Atomic Energy Agency (IAEA) reactor safety programs, and Research and Development Institute of Power Engineering (RDIPE) reports. The response to the PNL questions have not been edited or reviewed for technical consistency or accuracy by PNL staff or other US organizations, but are provided for use by the DOE community in the form they were received.

  1. BDDR, a new CEA technological and operating reactor database

    SciTech Connect

    Soldevilla, M.; Salmons, S.; Espinosa, B.

    2013-07-01

    The new application BDDR (Reactor database) has been developed at CEA in order to manage nuclear reactors technological and operating data. This application is a knowledge management tool which meets several internal needs: -) to facilitate scenario studies for any set of reactors, e.g. non-proliferation assessments; -) to make core physics studies easier, whatever the reactor design (PWR-Pressurized Water Reactor-, BWR-Boiling Water Reactor-, MAGNOX- Magnesium Oxide reactor-, CANDU - CANada Deuterium Uranium-, FBR - Fast Breeder Reactor -, etc.); -) to preserve the technological data of all reactors (past and present, power generating or experimental, naval propulsion,...) in a unique repository. Within the application database are enclosed location data and operating history data as well as a tree-like structure containing numerous technological data. These data address all kinds of reactors features and components. A few neutronics data are also included (neutrons fluxes). The BDDR application is based on open-source technologies and thin client/server architecture. The software architecture has been made flexible enough to allow for any change. (authors)

  2. ETUDE - European Trade Union Distance Education.

    ERIC Educational Resources Information Center

    Creanor, Linda; Walker, Steve

    2000-01-01

    Describes transnational distance learning activities among European trade union educators carried out as part of the European Trade Union Distance Education (ETUDE) project, supported by the European Commission. Highlights include the context of international trade union distance education; tutor training course; tutors' experiences; and…

  3. Centre National d'Etudes Spatiales

    NASA Astrophysics Data System (ADS)

    Murdin, P.

    2000-11-01

    The Centre National d'Etudes Spatiales (CNES) draws up, proposes and conducts France's space policy. Its role is to develop the uses of space, to meet the civilian and military needs of public bodies and of the scientific community, and to foster the development and dissemination of new applications, designed to create wealth and jobs....

  4. Reactor-specific spent fuel discharge projections: 1986 to 2020

    SciTech Connect

    Heeb, C.M.; Walling, R.C.; Purcell, W.L.

    1987-03-01

    The creation of five reactor-specific spent fuel data bases that contain information on the projected amounts of spent fuel to be discharged from US commercial nuclear reactors through the year 2020 is described. The data bases contain detailed spent-fuel information from existing, planned, and projected pressurized water reactors (PWR) and boiling water reactors (BWR). The projections are based on individual reactor information supplied by the US reactor owners. The basic information is adjusted to conform to Energy Information Agency (EIA) forecasts for nuclear installed capacity, generation, and spent fuel discharged. The EIA cases considered are: (1) No new orders with extended burnup, (2) No new orders with constant burnup, (3) Upper reference (which assumes extended burnup), (4) Upper reference with constant burnup, and (5) Lower reference (which assumes extended burnup). Detailed, by-reactor tables are provided for annual discharged amounts of spent fuel, for storage requirements assuming maximum-at-reactor storage, and for storage requirements assuming maximum-at-reactor plus intra-utility transshipment of spent fuel. 6 refs., 8 figs., 8 tabs.

  5. Reactor-specific spent fuel discharge projections, 1987-2020

    SciTech Connect

    Walling, R.C.; Heeb, C.M.; Purcell, W.L.

    1988-03-01

    The creation of five reactor-specific spent fuel data bases that contain information on the projected amounts of spent fuel to be discharged from U.S. commercial nuclear reactors through the year 2020 is described. The data bases contain detailed spent fuel information from existing, planned, and projected pressurized water reactors (PWR) and boiling water eactors (BWR), and one existing high temperature gas reactor (HTGR). The projections are based on individual reactor information supplied by the U.S. reactor owners. The basic information is adjusted to conform to Energy Information Administration (EIA) forecasts for nuclear installed capacity, generation, and spent fuel discharged. The EIA cases considered are: No New Orders (assumes increasing burnup), No New Orders with No Increased Burnup, Upper Reference (assumes increasing burnup), Upper Reference with No Increased Burnup, and Lower Reference (assumes increasing burnup). Detailed, by-reactor tables are provided for annual discharged amounts of spent fuel, for storage requirements assuming maximum at-reactor storage, and for storage requirements assuming maximum at-reactor storage plus intra-utility transshipment of spent fuel. 8 refs., 8 figs., 10 tabs.

  6. Analysis of nuclear reactor instability phenomena

    SciTech Connect

    Lahey, R.T. Jr.

    1993-01-01

    The phenomena known as density-wave instability often occurs in phase change systems, such as boiling water nuclear reactors (BWRS). Our current understanding of density-wave oscillations is in fairly good shape for linear phenomena (eg, the onset of instabilities) but is not very advanced for non-linear phenomena [Lahey and Podowski, 1989]. In particular, limit cycle and chaotic instability modes are not well understood in boiling systems such as current and advanced generation BWRs (eg, SBWR). In particular, the SBWR relies on natural circulation and is thus inherently prone to problems with density-wave instabilities. The purpose of this research is to develop a quantitative understanding of nonlinear nuclear-coupled density-wave instability phenomena in BWRS. This research builds on the work of Achard et al [1985] and Clausse et al [1991] who showed, respectively, that Hopf bifurcations and chaotic oscillations may occur in boiling systems.

  7. 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.

  8. Reactor-specific spent fuel discharge projections: 1985 to 2020

    SciTech Connect

    Heeb, C.M.; Libby, R.A.; Walling, R.C.; Purcell, W.L.

    1986-09-01

    The creation of four spent-fuel data bases that contain information on the projected amounts of spent fuel to be discharged from US commercial nuclear reactors through the year 2020 is described. The data bases contain detailed spent-fuel information from existing, planned, and projected pressurized water reactors (PWR) and boiling water reactors (BWR). The projections are based on individual reactor information supplied by the US reactor owners. The basic information is adjusted to conform to Energy Information Agency (EIA) forecasts for nuclear installed capacity, generation, and spent fuel discharged. The EIA cases considered are: (1) No New Orders with Extended Burnup, (2) No New Orders with Constant Burnup, (3) Middle Case with Extended Burnup, and (4) Middle Case with Constant Burnup. Detailed, by-reactor tables are provided for annual discharged amounts of spent fuel, for storage requirements assuming maximum-at-reactor storage, and for storage requirements assuming maximum-at-reactor plus intra-utility transshipment of spent fuel.

  9. 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.

  10. Results of Coupling a Thermal-Hydraulic Test Loop and University Research Reactor

    SciTech Connect

    Cetiner, Sacit M.; Edwards, Robert M.

    2002-07-01

    The coupling of a university thermal-hydraulic test loop and a simulated reactor is presented. The thermal-hydraulic test loop used in this work is a one-half height scaled version of General Electric's Simplified Boiling Water Reactor (SBWR). The digitally simulated reactor exploits modal neutron kinetics equations up to the first harmonic, and governing equations are not linearized. The preserved nonlinearity makes the simulated reactor behave more realistically, and eigenfunction expansion to the first order lets half of the core be represented independently. A series of experiments are performed with the hybrid system including simulated control rod reactivity insertion/withdrawal, cross-mode interaction, etc. The experimental results are compared with the theoretical expectations. (authors)

  11. The Development of Radiation Embrittlement Models for U. S. Power Reactor Pressure Vessel Steels

    SciTech Connect

    Wang, Jy-An John; Rao, Nageswara S; Konduri, Savanthi

    2007-01-01

    A new approach of utilizing information fusion technique is developed to predict the radiation embrittlement of reactor pressure vessel steels. The Charpy transition temperature shift data contained in the Power Reactor Embrittlement Database is used in this study. Six parameters {Cu, Ni, P, neutron fluence, irradiation time, and irradiation temperature {are used in the embrittlement prediction models. The results indicate that this new embrittlement predictor achieved reductions of about 49.5% and 52% in the uncertainties for plate and weld data, respectively, for pressurized water reactor and boiling water reactor data, compared with the Nuclear Regulatory Commission Regulatory Guide 1.99, Rev. 2. The implications of dose-rate effect and irradiation temperature effects for the development of radiation embrittlement models are also discussed.

  12. Problems in experimental and mathematical investigations of the accidental thermalhydraulic processes in RBMK nuclear reactors

    SciTech Connect

    Nigmatulin, B.I.; Tikhonenko, L.K.; Blinkov, V.N.

    1995-09-01

    In this paper the thermalhydraulic scheme and peculiarities of the boiling water graphite-moderated channel-type reactor RBMK are presented and discussed shortly. The essential for RBMK transient regimes, accidental situations and accompanying thermalhydraulic phenomena and processes are formulated. These data are presented in the form of cross reference matrix (version 1) for system computer codes verification. The paper includes qualitative analysis of the computer codes and integral facilities which have been used or can be used for RBMK transients and accidents investigations. The stability margins for RBMK-1000 and RBMK-1500 are shown.

  13. The Information Fusion Embrittlement Models for U.S. Power Reactor Pressure Vessel Steels

    SciTech Connect

    Wang, Jy-An John; Rao, Nageswara S; Konduri, Savanthi

    2007-01-01

    The complex nonlinear dependencies observed in typical reactor pressure vessel (RPV) material embrittlement data, as well as the inherent large uncertainties and scatter in the radiation embrittlement data, make prediction of radiation embrittlement a difficult task. Conventional statistical and deterministic approaches have only resulted in rather large uncertainties, in part because they do not fully exploit domain-specific mechanisms. The domain models built by researchers in the field, on the other hand, do not fully exploit the statistical and information content of the data. As evidenced in previous studies, it is unlikely that a single method, whether statistical, nonlinear, or domain model, will outperform all others. More generally, considering the complexity of the embrittlement prediction problem, it is highly unlikely that a single best method exists and is tractable, even in theory. In this paper, we propose to combine a number of complementary methods including domain models, neural networks, and nearest neighbor regressions (NNRs). Such a combination of methods has become possible because of recent developments in measurement-based optimal fusers in the area of information fusion. The information fusion technique is used to develop radiation embrittlement prediction models for reactor RPV steels from U.S. power reactors, including boiling water reactors and pressurized water reactors. The Charpy transition temperature-shift data is used as the primary index of RPV radiation embrittlement in this study. Six Cu, Ni, P, neutron fluence, irradiation time, and irradiation-parameters are used in the embrittlement prediction models. The results-temperature indicate that this new embrittlement predictor achieved reductions of about 49.5% and 52% in the uncertainties for plate and weld data, respectively, for pressurized water reactor and boiling water reactor data, compared with the Nuclear Regulatory Commission Regulatory Guide 1.99, Rev. 2. The

  14. 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.

  15. Reactor internal pump behavior during cavitation

    SciTech Connect

    Komita, Hideo; Usuki, Shouji; Fukuda, Shinichi )

    1989-11-01

    The characteristics of reactor internal pumps (RIPs) under cavitation conditions were experimentally evaluated in full scale with different water temperature parameters. The hydraulic performance and vibration behavior under cavitation conditions were clarified. An advanced boiling water reactor is equipped with RIPs for coolant recirculation. The RIP is a vertical, single-stage, mixed-flow pump that is mounted in the annular downcomer of the reactor pressure vessel. In general, when a pump operates under cavitation conditions, the pump total head decreases, and the appearance and collapse of vapor bubbles induce vibration and noise, causing damage to components. Various reports have been made on centrifugal pump behavior in cold water, but very few have discussed the characteristics of a mixed-flow pump like the RIP in hot water. Very few have measured impeller vibration in hot water, which is significantly influenced by cavitation. Therefore, it is difficult to precisely determine RIP behavior under cavitation conditions. Hydraulic performance and vibration behavior under cavitation conditions are experimentally clarified in this paper.

  16. Technology, Safety and Costs of Decommissioning Nuclear Reactors At Multiple-Reactor Stations

    SciTech Connect

    Wittenbrock, N. G.

    1982-01-01

    Safety and cost information is developed for the conceptual decommissioning of large (1175-MWe) pressurized water reactors (PWRs) and large (1155-MWe) boiling water reactors {BWRs) at multiple-reactor stations. Three decommissioning alternatives are studied: DECON (immediate decontamination), SAFSTOR (safe storage followed by deferred decontamination), and ENTOMB (entombment). Safety and costs of decommissioning are estimated by determining the impact of probable features of multiple-reactor-station operation that are considered to be unavailable at a single-reactor station, and applying these estimated impacts to the decommissioning costs and radiation doses estimated in previous PWR and BWR decommissioning studies. The multiple-reactor-station features analyzed are: the use of interim onsite nuclear waste storage with later removal to an offsite nuclear waste disposal facility, the use of permanent onsite nuclear waste disposal, the dedication of the site to nuclear power generation, and the provision of centralized services. Five scenarios for decommissioning reactors at a multiple-reactor station are investigated. The number of reactors on a site is assumed to be either four or ten; nuclear waste disposal is varied between immediate offsite disposal, interim onsite storage, and immediate onsite disposal. It is assumed that the decommissioned reactors are not replaced in one scenario but are replaced in the other scenarios. Centralized service facilities are provided in two scenarios but are not provided in the other three. Decommissioning of a PWR or a BWR at a multiple-reactor station probably will be less costly and result in lower radiation doses than decommissioning an identical reactor at a single-reactor station. Regardless of whether the light water reactor being decommissioned is at a single- or multiple-reactor station: • the estimated occupational radiation dose for decommissioning an LWR is lowest for SAFSTOR and highest for DECON • the estimated

  17. 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.

  18. Revisiting the structure of the anti-neoplastic glucans of Mycobacterium bovis Bacille Calmette-Guerin. Structural analysis of the extracellular and boiling water extract-derived glucans of the vaccine substrains.

    PubMed

    Dinadayala, Premkumar; Lemassu, Anne; Granovski, Pierre; Cérantola, Stéphane; Winter, Nathalie; Daffé, Mamadou

    2004-03-26

    The attenuated strain of Mycobacterium bovis Bacille Calmette-Guérin (BCG), used worldwide to prevent tuberculosis and leprosy, is also clinically used as an immunotherapeutic agent against superficial bladder cancer. An anti-tumor polysaccharide has been isolated from the boiling water extract of the Tice substrain of BCG and tentatively characterized as consisting primarily of repeating units of 6-linked-glucosyl residues. Mycobacterium tuberculosis and other mycobacterial species produce a glycogen-like alpha-glucan composed of repeating units of 4-linked glucosyl residues substituted at some 6 positions by short oligoglucosyl units that also exhibits an anti-tumor activity. Therefore, the impression prevails that mycobacteria synthesize different types of anti-neoplastic glucans or, alternatively, the BCG substrains are singular in producing a unique type of glucan that may confer to them their immunotherapeutic property. The present study addresses this question through the comparative analysis of alpha-glucans purified from the extracellular materials and boiling water extracts of three vaccine substrains. The polysaccharides were purified, and their structural features were established by mono- and two-dimensional NMR spectroscopy and matrix-assisted laser desorption/ionization time-of-flight mass spectrometry of the enzymatic and chemical degradation products of the purified compounds. The glucans isolated by the two methods from the three substrains of BCG were shown to exhibit identical structural features shared with the glycogen-like alpha-glucan of M. tuberculosis and other mycobacteria. Incidentally, we observed an occasional release of dextrans from Sephadex columns that may explain the reported occurrence of 6-substituted alpha-glucans in mycobacteria.

  19. 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.

  20. 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.

  1. A simplified model of aerosol removal by natural processes in reactor containments

    SciTech Connect

    Powers, D.A.; Washington, K.E.; Sprung, J.L.; Burson, S.B.

    1996-07-01

    Simplified formulae are developed for estimating the aerosol decontamination that can be achieved by natural processes in the containments of pressurized water reactors and in the drywells of boiling water reactors under severe accident conditions. These simplified formulae were derived by correlation of results of Monte Carlo uncertainty analyses of detailed models of aerosol behavior under accident conditions. Monte Carlo uncertainty analyses of decontamination by natural aerosol processes are reported for 1,000, 2,000, 3,000, and 4,000 MW(th) pressurized water reactors and for 1,500, 2,500, and 3,500 MW(th) boiling water reactors. Uncertainty distributions for the decontamination factors and decontamination coefficients as functions of time were developed in the Monte Carlo analyses by considering uncertainties in aerosol processes, material properties, reactor geometry and severe accident progression. Phenomenological uncertainties examined in this work included uncertainties in aerosol coagulation by gravitational collision, Brownian diffusion, turbulent diffusion and turbulent inertia. Uncertainties in aerosol deposition by gravitational settling, thermophoresis, diffusiophoresis, and turbulent diffusion were examined. Electrostatic charging of aerosol particles in severe accidents is discussed. Such charging could affect both the coagulation and deposition of aerosol particles. Electrostatic effects are not considered in most available models of aerosol behavior during severe accidents and cause uncertainties in predicted natural decontamination processes that could not be taken in to account in this work. Median (50%), 90 and 10% values of the uncertainty distributions for effective decontamination coefficients were correlated with time and reactor thermal power. These correlations constitute a simplified model that can be used to estimate the decontamination by natural aerosol processes at 3 levels of conservatism. Applications of the model are described.

  2. Reactor subchannel analysis -- Electric Power Research Institute perspective

    SciTech Connect

    Srikantiah, G.

    1995-12-01

    One of the basic objectives of subchannel flow simulation and analysis effort sponsored by the Electric Power Research Institute was the development of a computer code for subchannel analysis and its verification and validation for applications to reactor thermal margin evaluation under steady and transient conditions. A historical perspective is given of the development of specifications for a reactor core subchannel thermal-hydraulics analysis code for utility applications in the evaluation of reactor safety limits during normal operation and accident scenarios. The subchannel analysis capabilities of the VIPRE-01 code based on the homogeneous equilibrium with the algebraic slip model of two-phase flow are presented. The code, which received a safety evaluation report from the US Nuclear Regulatory Commission in 1986, is in wide use in the utility industry for fuel reload safety analysis, critical heat flux correlation development and testing, thermal margin analysis, and core thermal-hydraulic analysis. A considerable amount of work has been done during the past few years on the development of VIPRE-02, an advanced subchannel analysis code based on the two-fluid model of two-phase flow capable of simulating reactor cores, vessels, and internal structures. The functional specifications, development of VIPRE-02, and current applications for VIPRE-02, such as boiling water reactor mixed fuel core evaluation, are also discussed. Code is also used for PWR`s.

  3. Analysis of nuclear reactor instability phenomena. Progress report

    SciTech Connect

    Lahey, R.T. Jr.

    1993-03-01

    The phenomena known as density-wave instability often occurs in phase change systems, such as boiling water nuclear reactors (BWRS). Our current understanding of density-wave oscillations is in fairly good shape for linear phenomena (eg, the onset of instabilities) but is not very advanced for non-linear phenomena [Lahey and Podowski, 1989]. In particular, limit cycle and chaotic instability modes are not well understood in boiling systems such as current and advanced generation BWRs (eg, SBWR). In particular, the SBWR relies on natural circulation and is thus inherently prone to problems with density-wave instabilities. The purpose of this research is to develop a quantitative understanding of nonlinear nuclear-coupled density-wave instability phenomena in BWRS. This research builds on the work of Achard et al [1985] and Clausse et al [1991] who showed, respectively, that Hopf bifurcations and chaotic oscillations may occur in boiling systems.

  4. Fast Reactors

    NASA Astrophysics Data System (ADS)

    Esposito, S.; Pisanti, O.

    The following sections are included: * Elementary Considerations * The Integral Equation to the Neutron Distribution * The Critical Size for a Fast Reactor * Supercritical Reactors * Problems and Exercises

  5. Tritium production, releases and population doses at nuclear power reactors

    SciTech Connect

    Peterson, H.T.; Baker, D.A.

    1985-09-01

    Tritium is produced in light-water-cooled reactors as a product of ternary fission and by nuclear reactions with the coolant and with neutron-absorbing materials used for reactor control. Pressurized water reactors (PWRs) have greater amounts of tritium produced in or released into the coolant than boiling water reactors (BWRs). Consequently, tritium releases to the environment from PWRs (29 GBq/MW(e)-y (0.78 Ci/MW(e)-y)/sup 0/ are about 6 1/2 times greater than from BWRs (4.4 GBq/MW(e)-y (0.12 Ci/MW(e)-y)/sup 0/. Most of the tritium released from PWRs appears in the liquid effluent (about 85%), whereas 75% of the tritium released from BWRs is as airborne effluents. Radiation doses from these tritium releases are small; the average site collective (population) dose in 1981 was less than 0.002 person-sieverts per year (0.2 person-rem/ year). The total collective dose from all tritium releases was 0.08 personsieverts (8 person-rem).

  6. System and method for determining coolant level and flow velocity in a nuclear reactor

    DOEpatents

    Brisson, Bruce William; Morris, William Guy; Zheng, Danian; Monk, David James; Fang, Biao; Surman, Cheryl Margaret; Anderson, David Deloyd

    2013-09-10

    A boiling water reactor includes a reactor pressure vessel having a feedwater inlet for the introduction of recycled steam condensate and/or makeup coolant into the vessel, and a steam outlet for the discharge of produced steam for appropriate work. A fuel core is located within a lower area of the pressure vessel. The fuel core is surrounded by a core shroud spaced inward from the wall of the pressure vessel to provide an annular downcomer forming a coolant flow path between the vessel wall and the core shroud. A probe system that includes a combination of conductivity/resistivity probes and/or one or more time-domain reflectometer (TDR) probes is at least partially located within the downcomer. The probe system measures the coolant level and flow velocity within the downcomer.

  7. A flooding induced station blackout analysis for a pressurized water reactor using the RISMC toolkit

    DOE PAGES

    Mandelli, Diego; Prescott, Steven; Smith, Curtis; ...

    2015-05-17

    In this paper we evaluate the impact of a power uprate on a pressurized water reactor (PWR) for a tsunami-induced flooding test case. This analysis is performed using the RISMC toolkit: the RELAP-7 and RAVEN codes. RELAP-7 is the new generation of system analysis codes that is responsible for simulating the thermal-hydraulic dynamics of PWR and boiling water reactor systems. RAVEN has two capabilities: to act as a controller of the RELAP-7 simulation (e.g., component/system activation) and to perform statistical analyses. In our case, the simulation of the flooding is performed by using an advanced smooth particle hydrodynamics code calledmore » NEUTRINO. The obtained results allow the user to investigate and quantify the impact of timing and sequencing of events on system safety. The impact of power uprate is determined in terms of both core damage probability and safety margins.« less

  8. A flooding induced station blackout analysis for a pressurized water reactor using the RISMC toolkit

    SciTech Connect

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

    2015-05-17

    In this paper we evaluate the impact of a power uprate on a pressurized water reactor (PWR) for a tsunami-induced flooding test case. This analysis is performed using the RISMC toolkit: the RELAP-7 and RAVEN codes. RELAP-7 is the new generation of system analysis codes that is responsible for simulating the thermal-hydraulic dynamics of PWR and boiling water reactor systems. RAVEN has two capabilities: to act as a controller of the RELAP-7 simulation (e.g., component/system activation) and to perform statistical analyses. In our case, the simulation of the flooding is performed by using an advanced smooth particle hydrodynamics code called NEUTRINO. The obtained results allow the user to investigate and quantify the impact of timing and sequencing of events on system safety. The impact of power uprate is determined in terms of both core damage probability and safety margins.

  9. Impact of Reactor Operating Parameters on Cask Reactivity in BWR Burnup Credit

    SciTech Connect

    Ilas, Germina; Betzler, Benjamin R; Ade, Brian J

    2017-01-01

    This paper discusses the effect of reactor operating parameters used in fuel depletion calculations on spent fuel cask reactivity, with relevance for boiling-water reactor (BWR) burnup credit (BUC) applications. Assessments that used generic BWR fuel assembly and spent fuel cask configurations are presented. The considered operating parameters, which were independently varied in the depletion simulations for the assembly, included fuel temperature, bypass water density, specific power, and operating history. Different operating history scenarios were considered for the assembly depletion to determine the effect of relative power distribution during the irradiation cycles, as well as the downtime between cycles. Depletion, decay, and criticality simulations were performed using computer codes and associated nuclear data within the SCALE code system. Results quantifying the dependence of cask reactivity on the assembly depletion parameters are presented herein.

  10. NEUTRONIC REACTOR

    DOEpatents

    Fermi, E.; Zinn, W.H.; Anderson, H.L.

    1958-09-16

    Means are presenied for increasing the reproduction ratio of a gaphite- moderated neutronic reactor by diminishing the neutron loss due to absorption or capture by gaseous impurities within the reactor. This means comprised of a fluid-tight casing or envelope completely enclosing the reactor and provided with a valve through which the casing, and thereby the reactor, may be evacuated of atmospheric air.

  11. Review of the proposed materials of construction for the SBWR and AP600 advanced reactors

    SciTech Connect

    Diercks, D.R.; Shack, W.J.; Chung, H.M.; Kassner, T.F.

    1994-06-01

    Two advanced light water reactor (LWR) concepts, namely the General Electric Simplified Boiling Water Reactor (SBWR) and the Westinghouse Advanced Passive 600 MWe Reactor (AP600), were reviewed in detail by Argonne National Laboratory. The objectives of these reviews were to (a) evaluate proposed advanced-reactor designs and the materials of construction for the safety systems, (b) identify all aging and environmentally related degradation mechanisms for the materials of construction, and (c) evaluate from the safety viewpoint the suitability of the proposed materials for the design application. Safety-related systems selected for review for these two LWRs included (a) reactor pressure vessel, (b) control rod drive system and reactor internals, (c) coolant pressure boundary, (d) engineered safety systems, (e) steam generators (AP600 only), (f) turbines, and (g) fuel storage and handling system. In addition, the use of cobalt-based alloys in these plants was reviewed. The selected materials for both reactors were generally sound, and no major selection errors were found. It was apparent that considerable thought had been given to the materials selection process, making use of lessons learned from previous LWR experience. The review resulted in the suggestion of alternate an possibly better materials choices in a number of cases, and several potential problem areas have been cited.

  12. Adaptation of Crack Growth Detection Techniques to US Material Test Reactors

    SciTech Connect

    A. Joseph Palmer; Sebastien P. Teysseyre; Kurt L. Davis; Joy L. Rempe; Gordon Kohse; Yakov Ostrovsky; David M. Carpenter

    2014-04-01

    A key component in evaluating the ability of Light Water Reactors to operate beyond 60 years is characterizing the degradation of materials exposed to radiation and various water chemistries. Of particular concern is the response of reactor materials to Irradiation Assisted Stress Corrosion Cracking (IASCC). Some materials testing reactors (MTRs) outside the U.S., such as the Halden Boiling Water Reactor (HBWR), have deployed a technique to measure crack growth propagation during irradiation. This technique incorporates a compact loading mechanism to stress the specimen during irradiation. A crack in the specimen is monitored using the Direct Current Potential Drop (DCPD) method. A project is underway to develop and demonstrate the performance of a similar type of test rig for use in U.S. MTRs. The first year of this three year project was devoted to designing, analyzing, fabricating, and bench top testing a mechanism capable of applying a controlled stress to specimens while they are irradiated in a pressurized water loop (simulating PWR reactor conditions). During the second year, the mechanism will be tested in autoclaves containing high pressure, high temperature water with representative water chemistries. In addition, necessary documentation and safety reviews for testing in a reactor environment will be completed. In the third year, the assembly will be tested in the Massachusetts Institute of Technology Reactor (MITR) and Post Irradiation Examinations (PIE) will be performed.

  13. Simulation of (16)O (n, p) (16)N reaction rate and nitrogen-16 inventory in a high performance light water reactor with one pass core.

    PubMed

    Kebwaro, Jeremiah Monari; Zhao, Yaolin; He, Chaohui

    2014-12-01

    The rate of activation of the isotope (16)O to (16)N in a typical HPLWR one pass concept was calculated using MCNP code. A mathematical model was used to track the inventory of the radioisotope (16)N in a unit mass of coolant traversing the system. The water leaving the moderator channels has the highest activity in the circuit, but due to interaction with fresh coolant at the lower plenum, the activity is downscaled. The calculated core exit activity is higher than values reported in literature for commercial boiling water reactors.

  14. Thermoacoustic Thermometry for Nuclear Reactor Monitoring

    SciTech Connect

    James A. Smith; Dale K. Kotter; Steven L. Garrett; Randall A. Ali

    2013-06-01

    On Friday, March 11, 2011, at 2:46pm (Japan Standard Trme), the Tohoku region on the east coast of northern Japan experi­enced what would become known as the largest earthquake in the country's history at magnitude 9.0 on the Richter scale. The Fukushima Daiichi nuclear power plant suffered exten­sive and irreversible damage. Six operating units were at the site, each with a boiling water reactor. When the earthquake struck, three of the six reactors were operating and the others were in a periodic inspection outage phase. In one reactor, all of the fuel had been relocated to a spent fuel pool in the reactor building. The seismic acceleration caused by the earthquake brought the three operating units to an automatic shutdown. Since there was damage to the power transmission lines, the emergency diesel generators (EDG) were automat­ically started to ensure continued cooling of the reactors and spent fuel pools. The situation was under control until the tsunami hit about forty-five minutes later with a maximum wave height of approximately 15 meters, which was three times taller than the sea wall of 5m. The influx of water submerged the EDGs, the electrical switchgear, and dc batteries, resulting in the total loss of power to five of the six reactors. The flooding also resulted in the loss of instrumentation that would have other­ wise been used to monitor and control the emergency. The ugly aftermath included high radiation exposure to operators at the nuclear power plants and early contamina­tion of food supplies and water within several restricted areas in Japan, where high radiation levels have rendered them un­safe for human habitation. While the rest of the story will remain a tragic history, it is this part of the series of unfortunate events that has inspired our research. It has indubitably highlighted the need for a novel sensor and instrumentation system that can withstand similar or worse conditions to avoid future catastrophe and assume damage

  15. 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.

  16. 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.

  17. 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.

  18. 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.

  19. Use of probabilistic inversion to model qualitative expert input when selecting a new nuclear reactor technology

    NASA Astrophysics Data System (ADS)

    Merritt, Charles R., Jr.

    Complex investment decisions by corporate executives often require the comparison of dissimilar attributes and competing technologies. A technique to evaluate qualitative input from experts using a Multi-Criteria Decision Method (MCDM) is described to select a new reactor technology for a merchant nuclear generator. The high capital cost, risks from design, licensing and construction, reactor safety and security considerations are some of the diverse considerations when choosing a reactor design. Three next generation reactor technologies are examined: the Advanced Pressurized-1000 (AP-1000) from Westinghouse, Economic Simplified Boiling Water Reactor (ESBWR) from General Electric, and the U.S. Evolutionary Power Reactor (U.S. EPR) from AREVA. Recent developments in MCDM and decision support systems are described. The uncertainty inherent in experts' opinions for the attribute weighting in the MCDM is modeled through the use of probabilistic inversion. In probabilistic inversion, a function is inverted into a random variable within a defined range. Once the distribution is created, random samples based on the distribution are used to perform a sensitivity analysis on the decision results to verify the "strength" of the results. The decision results for the pool of experts identified the U.S. EPR as the optimal choice.

  20. OECD NEA Benchmark Database of Spent Nuclear Fuel Isotopic Compositions for World Reactor Designs

    SciTech Connect

    Gauld, Ian C; Sly, Nicholas C; Michel-Sendis, Franco

    2014-01-01

    Experimental data on the isotopic concentrations in irradiated nuclear fuel represent one of the primary methods for validating computational methods and nuclear data used for reactor and spent fuel depletion simulations that support nuclear fuel cycle safety and safeguards programs. Measurement data have previously not been available to users in a centralized or searchable format, and the majority of accessible information has been, for the most part, limited to light-water-reactor designs. This paper describes a recent initiative to compile spent fuel benchmark data for additional reactor designs used throughout the world that can be used to validate computer model simulations that support nuclear energy and nuclear safeguards missions. Experimental benchmark data have been expanded to include VVER-440, VVER-1000, RBMK, graphite moderated MAGNOX, gas cooled AGR, and several heavy-water moderated CANDU reactor designs. Additional experimental data for pressurized light water and boiling water reactor fuels has also been compiled for modern assembly designs and more extensive isotopic measurements. These data are being compiled and uploaded to a recently revised structured and searchable database, SFCOMPO, to provide the nuclear analysis community with a centrally-accessible resource of spent fuel compositions that can be used to benchmark computer codes, models, and nuclear data. The current version of SFCOMPO contains data for eight reactor designs, 20 fuel assembly designs, more than 550 spent fuel samples, and measured isotopic data for about 80 nuclides.

  1. BOILING REACTORS

    DOEpatents

    Untermyer, S.

    1962-04-10

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

  2. NEUTRONIC REACTOR

    DOEpatents

    Daniels, F.

    1959-10-27

    A reactor in which at least a portion of the moderator is in the form of movable refractory balls is described. In addition to their moderating capacity, these balls may serve as carriers for fissionable material or fertile material, or may serve in a coolant capacity to remove heat from the reactor. A pneumatic system is used to circulate the balls through the reactor.

  3. Generic analyses for evaluation of low Charpy upper-shelf energy effects on safety margins against fracture of reactor pressure vessel materials

    SciTech Connect

    Dickson, T.L.

    1993-07-01

    Appendix G to 10 CFR Part 50 requires that reactor pressure vessel beltline material maintain Charpy upper-shelf energies of no less than 50 ft-lb during the plant operating life, unless it is demonstrated in a manner approved by the Nuclear Regulatory Commission (NRC), that lower values of Charpy upper-shelf energy provide margins of safety against fracture equivalent to those in Appendix G to Section XI of the ASME Code. Analyses based on acceptance criteria and analysis methods adopted in the ASME Code Case N-512 are described herein. Additional information on material properties was provided by the NRC, Office of Nuclear Regulatory Research, Materials Engineering Branch. These cases, specified by the NRC, represent generic applications to boiling water reactor and pressurized water reactor vessels. This report is designated as HSST Report No. 140.

  4. CONVECTION REACTOR

    DOEpatents

    Hammond, R.P.; King, L.D.P.

    1960-03-22

    An homogeneous nuclear power reactor utilizing convection circulation of the liquid fuel is proposed. The reactor has an internal heat exchanger looated in the same pressure vessel as the critical assembly, thereby eliminating necessity for handling the hot liquid fuel outside the reactor pressure vessel during normal operation. The liquid fuel used in this reactor eliminates the necessity for extensive radiolytic gas rocombination apparatus, and the reactor is resiliently pressurized and, without any movable mechanical apparatus, automatically regulates itself to the condition of criticality during moderate variations in temperature snd pressure and shuts itself down as the pressure exceeds a predetermined safe operating value.

  5. Research reactors

    SciTech Connect

    Tonneson, L.C.; Fox, G.J.

    1996-04-01

    There are currently 284 research reactors in operation, and 12 under construction around the world. Of the operating reactors, nearly two-thirds are used exclusively for research, and the rest for a variety of purposes, including training, testing, and critical assembly. For more than 50 years, research reactor programs have contributed greatly to the scientific and educational communities. Today, six of the world`s research reactors are being shut down, three of which are in the USA. With government budget constraints and the growing proliferation concerns surrounding the use of highly enriched uranium in some of these reactors, the future of nuclear research could be impacted.

  6. SCALE-4 Analysis of LaSalle Unit 1 BWR Commercial Reactor Critical Configurations

    SciTech Connect

    Gauld, I.C.

    2000-03-01

    Five commercial reactor criticals (CRCs) for the LaSalle Unit 1 boiling-water reactor have been analyzed using KENO V.a, the Monte Carlo criticality code of the SCALE 4 code system. The irradiated fuel assembly isotopics for the criticality analyses were provided by the Waste Package Design team at the Yucca Mountain Project in the US, who performed the depletion calculations using the SAS2H sequence of SCALE 4. The reactor critical measurements involved two beginning-of-cycle and three middle-of-cycle configurations. The CRCs involved relatively low-cycle burnups, and therefore contained a relatively high gadolinium poison content in the reactor assemblies. This report summarizes the data and methods used in analyzing the critical configurations and assesses the sensitivity of the results to some of the modeling approximations used to represent the gadolinium poison distribution within the assemblies. The KENO V.a calculations, performed using the SCALE 44GROUPNDF5 ENDF/B-V cross-section library, yield predicted k{sub eff} values within about 1% {Delta}k/k relative to reactor measurements for the five CRCs using general 8-pin and 9-pin heterogeneous gadolinium poison pin assembly models.

  7. SCALE-4 Analysis of LaSalle Unit 1 BWR Commercial Reactor Critical Configuration

    SciTech Connect

    Gauld, I.C.

    2000-03-16

    Five commercial reactor criticals (CRCs) for the LaSalle Unit 1 boiling-water reactor have been analyzed using KENO V.a, the Monte Carlo criticality code of the SCALE 4 code system. The irradiated fuel assembly isotopics for the criticality analyses were provided by the Waste Package Design team at the Yucca Mountain Project in the United States, who performed the depletion calculations using the SAS2H sequence of SCALE 4. The reactor critical measurements involved two beginning-of-cycle and three middle-of-cycle configurations. The CRCs involved relatively low-cycle burnups, and therefore contained a relatively high gadolinium poison content in the reactor assemblies. This report summarizes the data and methods used in analyzing the critical configurations and assesses the sensitivity of the results to some of the modeling approximations used to represent the gadolinium poison distribution within the assemblies. The KENO V.a calculations, performed using the SCALE 44GROUPNDF5 ENDF/B-V cross-section library, yield predicted k{sub eff} values within about 1% {Delta}k/k relative to reactor measurements for the five CRCs using general 8-pin and 9-pin heterogeneous gadolinium poison pin assembly models.

  8. A procedure for evaluating residual life of major components in light water reactors

    SciTech Connect

    Uchida, S.; Fujimori, H.; Ibe, E.; Kuniya, J.; Hayashi, M.; Fuse, M.; Yamauchi, K.

    1995-12-31

    A computer program for evaluating residual life of major components in boiling water reactors is proposed. It divides the stress corrosion cracking process into two stages; a probabilistic crack generation stage and a deterministic crack propagation one. The minimum period of the crack generation stage is evaluated assuming an exponential distribution of the stage. The crack propagation rate is calculated by the slip-dissolution/film-rupture model. The neutron flux and fluence dependence of the neutron radiation effects on material properties was evaluated by using theoretical models of radiation damage. The computer program works on an engineering work station. Evaluated results are displayed as a map of the residual life, or as graphs of crack length evolution.

  9. NEUTRONIC REACTOR

    DOEpatents

    Fraas, A.P.; Mills, C.B.

    1961-11-21

    A neutronic reactor in which neutron moderation is achieved primarily in its reflector is described. The reactor structure consists of a cylindrical central "island" of moderator and a spherical moderating reflector spaced therefrom, thereby providing an annular space. An essentially unmoderated liquid fuel is continuously passed through the annular space and undergoes fission while contained therein. The reactor, because of its small size, is particularly adapted for propulsion uses, including the propulsion of aircraft. (AEC)

  10. REACTOR COOLING

    DOEpatents

    Quackenbush, C.F.

    1959-09-29

    A nuclear reactor with provisions for selectively cooling the fuel elements is described. The reactor has a plurality of tubes extending throughout. Cylindrical fuel elements are disposed within the tubes and the coolant flows through the tubes and around the fuel elements. The fuel elements within the central portion of the reactor are provided with roughened surfaces of material. The fuel elements in the end portions of the tubes within the reactor are provlded with low conduction jackets and the fuel elements in the region between the central portion and the end portions are provided with smooth surfaces of high heat conduction material.

  11. 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.

  12. NEUTRONIC REACTOR

    DOEpatents

    Wigner, E.P.

    1958-04-22

    A nuclear reactor for isotope production is described. This reactor is designed to provide a maximum thermal neutron flux in a region adjacent to the periphery of the reactor rather than in the center of the reactor. The core of the reactor is generally centrally located with respect tn a surrounding first reflector, constructed of beryllium. The beryllium reflector is surrounded by a second reflector, constructed of graphite, which, in tune, is surrounded by a conventional thermal shield. Water is circulated through the core and the reflector and functions both as a moderator and a coolant. In order to produce a greatsr maximum thermal neutron flux adjacent to the periphery of the reactor rather than in the core, the reactor is designed so tbat the ratio of neutron scattering cross section to neutron absorption cross section averaged over all of the materials in the reflector is approximately twice the ratio of neutron scattering cross section to neutron absorption cross section averaged over all of the material of the core of the reactor.

  13. Reactor building

    SciTech Connect

    Hista, J. C.

    1984-09-18

    Reactor building comprising a vessel shaft anchored in a slab which is peripherally locked. This reactor building comprises a confinement enclosure within which are positioned internal structures constituted by an internal structure floor, a vessel shaft, a slab being positioned between the general floor and the internal structure floor, the vesse

  14. NEUTRONIC REACTOR

    DOEpatents

    Metcalf, H.E.; Johnson, H.W.

    1961-04-01

    BS>A nuclear reactor incorporating fuel rods passing through a moderator and including tubes of a material of higher Thermal conductivity than the fuel in contact with the fuel is described. The tubes extend beyond the active portion of the reactor into contant with a fiuld coolant.

  15. Studies on the closed-loop digital control of multi-modular reactors. Final report

    SciTech Connect

    Bernard, J.A.; Henry, A.F.; Lanning, D.D.; Meyer, J.E.

    1992-11-01

    This report describes the theoretical development and the evaluation via both experiment and simulation of digital methods for the closed-loop control of power, temperature, and steam generator level in multi-modular reactors. The major conclusion of the research reported here is that the technology is currently available to automate many aspects of the operation of multi-modular plants. This will in turn minimize the number of required personnel and thus contain both operating and personnel costs, allow each module to be operated at a different power level thereby staggering the times at which refuelings would be needed, and maintain the competitiveness of US industry relative to foreign vendors who are developing and applying advanced control concepts. The technology described in this report is appropriate to the proposed multi-modular reactor designs and to present-generation pressurized water reactors. Its extension to boiling water reactors is possible provided that the commitment is made to create a real-time model of a BWR. The work reported here was performed by the Massachusetts Institute of Technology (MIT) under contract to the Oak Ridge National Laboratory (ORNL) and to the United States Department of Energy (Division of Industry and University Programs, Contract No. DE-FG07-90ER12930.)

  16. Studies on the closed-loop digital control of multi-modular reactors

    SciTech Connect

    Bernard, J.A. . Nuclear Reactor Lab.); Henry, A.F.; Lanning, D.D.; Meyer, J.E. . Dept. of Nuclear Engineering)

    1992-11-01

    This report describes the theoretical development and the evaluation via both experiment and simulation of digital methods for the closed-loop control of power, temperature, and steam generator level in multi-modular reactors. The major conclusion of the research reported here is that the technology is currently available to automate many aspects of the operation of multi-modular plants. This will in turn minimize the number of required personnel and thus contain both operating and personnel costs, allow each module to be operated at a different power level thereby staggering the times at which refuelings would be needed, and maintain the competitiveness of US industry relative to foreign vendors who are developing and applying advanced control concepts. The technology described in this report is appropriate to the proposed multi-modular reactor designs and to present-generation pressurized water reactors. Its extension to boiling water reactors is possible provided that the commitment is made to create a real-time model of a BWR. The work reported here was performed by the Massachusetts Institute of Technology (MIT) under contract to the Oak Ridge National Laboratory (ORNL) and to the United States Department of Energy (Division of Industry and University Programs, Contract No. DE-FG07-90ER12930.)

  17. 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.

  18. 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.

  19. 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.

  20. 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.

  1. Compact Reactor

    NASA Astrophysics Data System (ADS)

    Williams, Pharis E.

    2007-01-01

    Weyl's Gauge Principle of 1929 has been used to establish Weyl's Quantum Principle (WQP) that requires that the Weyl scale factor should be unity. It has been shown that the WQP requires the following: quantum mechanics must be used to determine system states; the electrostatic potential must be non-singular and quantified; interactions between particles with different electric charges (i.e. electron and proton) do not obey Newton's Third Law at sub-nuclear separations, and nuclear particles may be much different than expected using the standard model. The above WQP requirements lead to a potential fusion reactor wherein deuterium nuclei are preferentially fused into helium nuclei. Because the deuterium nuclei are preferentially fused into helium nuclei at temperatures and energies lower than specified by the standard model there is no harmful radiation as a byproduct of this fusion process. Therefore, a reactor using this reaction does not need any shielding to contain such radiation. The energy released from each reaction and the absence of shielding makes the deuterium-plus-deuterium-to-helium (DDH) reactor very compact when compared to other reactors, both fission and fusion types. Moreover, the potential energy output per reactor weight and the absence of harmful radiation makes the DDH reactor an ideal candidate for space power. The logic is summarized by which the WQP requires the above conditions that make the prediction of DDH possible. The details of the DDH reaction will be presented along with the specifics of why the DDH reactor may be made to cause two deuterium nuclei to preferentially fuse to a helium nucleus. The presentation will also indicate the calculations needed to predict the reactor temperature as a function of fuel loading, reactor size, and desired output and will include the progress achieved to date.

  2. Compact Reactor

    SciTech Connect

    Williams, Pharis E.

    2007-01-30

    Weyl's Gauge Principle of 1929 has been used to establish Weyl's Quantum Principle (WQP) that requires that the Weyl scale factor should be unity. It has been shown that the WQP requires the following: quantum mechanics must be used to determine system states; the electrostatic potential must be non-singular and quantified; interactions between particles with different electric charges (i.e. electron and proton) do not obey Newton's Third Law at sub-nuclear separations, and nuclear particles may be much different than expected using the standard model. The above WQP requirements lead to a potential fusion reactor wherein deuterium nuclei are preferentially fused into helium nuclei. Because the deuterium nuclei are preferentially fused into helium nuclei at temperatures and energies lower than specified by the standard model there is no harmful radiation as a byproduct of this fusion process. Therefore, a reactor using this reaction does not need any shielding to contain such radiation. The energy released from each reaction and the absence of shielding makes the deuterium-plus-deuterium-to-helium (DDH) reactor very compact when compared to other reactors, both fission and fusion types. Moreover, the potential energy output per reactor weight and the absence of harmful radiation makes the DDH reactor an ideal candidate for space power. The logic is summarized by which the WQP requires the above conditions that make the prediction of DDH possible. The details of the DDH reaction will be presented along with the specifics of why the DDH reactor may be made to cause two deuterium nuclei to preferentially fuse to a helium nucleus. The presentation will also indicate the calculations needed to predict the reactor temperature as a function of fuel loading, reactor size, and desired output and will include the progress achieved to date.

  3. NUCLEAR REACTOR

    DOEpatents

    Moore, R.V.; Bowen, J.H.; Dent, K.H.

    1958-12-01

    A heterogeneous, natural uranium fueled, solid moderated, gas cooled reactor is described, in which the fuel elements are in the form of elongated rods and are dlsposed within vertical coolant channels ln the moderator symmetrically arranged as a regular lattice in groups. This reactor employs control rods which operate in vertical channels in the moderator so that each control rod is centered in one of the fuel element groups. The reactor is enclosed in a pressure vessel which ls provided with access holes at the top to facilitate loading and unloadlng of the fuel elements, control rods and control rod driving devices.

  4. NEUTRONIC REACTOR

    DOEpatents

    Fermi, E.

    1960-04-01

    A nuclear reactor is described consisting of blocks of graphite arranged in layers, natural uranium bodies disposed in holes in alternate layers of graphite blocks, and coolant tubes disposed in the layers of graphite blocks which do not contain uranium.

  5. 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.

  6. NEUTRONIC REACTOR

    DOEpatents

    Anderson, H.L.

    1960-09-20

    A nuclear reactor is described comprising fissionable material dispersed in graphite blocks, helium filling the voids of the blocks and the spaces therebetween, and means other than the helium in thermal conductive contact with the graphite for removing heat.

  7. Reactor apparatus

    DOEpatents

    Echtler, J. Paul

    1981-01-01

    A reactor apparatus for hydrocracking a polynuclear aromatic hydrocarbonaceous feedstock to produce lighter hydrocarbon fuels by contacting the hydrocarbonaceous feedstock with hydrogen in the presence of a molten metal halide catalyst.

  8. Chemical Reactors.

    ERIC Educational Resources Information Center

    Kenney, C. N.

    1980-01-01

    Describes a course, including content, reading list, and presentation on chemical reactors at Cambridge University, England. A brief comparison of chemical engineering education between the United States and England is also given. (JN)

  9. NEUTRONIC REACTOR

    DOEpatents

    Hurwitz, H. Jr.; Brooks, H.; Mannal, C.; Payne, J.H.; Luebke, E.A.

    1959-03-24

    A reactor of the heterogeneous, liquid cooled type is described. This reactor is comprised of a central region of a plurality of vertically disposed elongated tubes surrounded by a region of moderator material. The central region is comprised of a central core surrounded by a reflector region which is surrounded by a fast neutron absorber region, which in turn is surrounded by a slow neutron absorber region. Liquid sodium is used as the primary coolant and circulates through the core which contains the fuel elements. Control of the reactor is accomplished by varying the ability of the reflector region to reflect neutrons back into the core of the reactor. For this purpose the reflector is comprised of moderator and control elements having varying effects on reactivity, the control elements being arranged and actuated by groups to give regulation, shim, and safety control.

  10. Reactor Engineering

    NASA Astrophysics Data System (ADS)

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

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

  11. NUCLEAR REACTOR

    DOEpatents

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

    1958-01-21

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

  12. NUCLEAR REACTOR

    DOEpatents

    Sherman, J.; Sharbaugh, J.E.; Fauth, W.L. Jr.; Palladino, N.J.; DeHuff, P.G.

    1962-10-23

    A nuclear reactor incorporating seed and blanket assemblies is designed. Means are provided for obtaining samples of the coolant from the blanket assemblies and for varying the flow of coolant through the blanket assemblies. (AEC)

  13. REACTOR SHIELD

    DOEpatents

    Wigner, E.P.; Ohlinger, L.E.; Young, G.J.; Weinberg, A.M.

    1959-02-17

    Radiation shield construction is described for a nuclear reactor. The shield is comprised of a plurality of steel plates arranged in parallel spaced relationship within a peripheral shell. Reactor coolant inlet tubes extend at right angles through the plates and baffles are arranged between the plates at right angles thereto and extend between the tubes to create a series of zigzag channels between the plates for the circulation of coolant fluid through the shield. The shield may be divided into two main sections; an inner section adjacent the reactor container and an outer section spaced therefrom. Coolant through the first section may be circulated at a faster rate than coolant circulated through the outer section since the area closest to the reactor container is at a higher temperature and is more radioactive. The two sections may have separate cooling systems to prevent the coolant in the outer section from mixing with the more contaminated coolant in the inner section.

  14. NEUTRONIC REACTORS

    DOEpatents

    Vernon, H.C.

    1959-01-13

    A neutronic reactor of the heterogeneous, fluid cooled tvpe is described. The reactor is comprised of a pressure vessel containing the moderator and a plurality of vertically disposed channels extending in spaced relationship through the moderator. Fissionable fuel material is placed within the channels in spaced relationship thereto to permit circulation of the coolant fluid. Separate means are provided for cooling the moderator and for circulating a fluid coolant thru the channel elements to cool the fuel material.

  15. NUCLEAR REACTOR

    DOEpatents

    Anderson, C.R.

    1962-07-24

    A fluidized bed nuclear reactor and a method of operating such a reactor are described. In the design means are provided for flowing a liquid moderator upwardly through the center of a bed of pellets of a nentron-fissionable material at such a rate as to obtain particulate fluidization while constraining the lower pontion of the bed into a conical shape. A smooth circulation of particles rising in the center and falling at the outside of the bed is thereby established. (AEC)

  16. NUCLEAR REACTOR

    DOEpatents

    Breden, C.R.; Dietrich, J.R.

    1961-06-20

    A water-soluble non-volatile poison may be introduced into a reactor to nullify excess reactivity. The poison is removed by passing a side stream of the water containing the soluble poison to an evaporation chamber. The vapor phase is returned to the reactor to decrease the concentration of soluble poison and the liquid phase is returned to increase the concentration of soluble poison.

  17. NUCLEAR REACTOR

    DOEpatents

    Grebe, J.J.

    1959-07-14

    High temperature reactors which are uniquely adapted to serve as the heat source for nuclear pcwered rockets are described. The reactor is comprised essentially of an outer tubular heat resistant casing which provides the main coolant passageway to and away from the reactor core within the casing and in which the working fluid is preferably hydrogen or helium gas which is permitted to vaporize from a liquid storage tank. The reactor core has a generally spherical shape formed entirely of an active material comprised of fissile material and a moderator material which serves as a diluent. The active material is fabricated as a gas permeable porous material and is interlaced in a random manner with very small inter-connecting bores or capillary tubes through which the coolant gas may flow. The entire reactor is divided into successive sections along the direction of the temperature gradient or coolant flow, each section utilizing materials of construction which are most advantageous from a nuclear standpoint and which at the same time can withstand the operating temperature of that particular zone. This design results in a nuclear reactor characterized simultaneously by a minimum critiral size and mass and by the ability to heat a working fluid to an extremely high temperature.

  18. Environmentally assisted cracking in light-water reactors: Semi-annual report, January--June 1997. Volume 24

    SciTech Connect

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

    1998-04-01

    This report summarizes work performed by Argonne National Laboratory on fatigue and environmentally assisted cracking (EAC) in light water reactors from January 1997 to June 1997. 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 Types 304 and 304L 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 is equally effective in decreasing fatigue life. Slow-strain-rate-tensile tests were conducted in simulated boiling water reactor (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 several heats of Alloys 600 and 690 in low-DO, simulated pressurized water reactor environments.

  19. Occupational radiation exposure at commercial nuclear power reactors and other facilities 1995: Twenty-eighth annual report. Volume 17

    SciTech Connect

    Thomas, M.L.; Hagemeyer, D.

    1997-01-01

    This report summarizes the occupational exposure data that are maintained in the US Nuclear Regulatory Commission`s (NRC) Radiation Exposure Information and Reporting System (REIRS). The bulk of the information contained in the report was compiled from the 1995 annual reports submitted by six of the seven categories of NRC licensees subject to the reporting requirements of 10 CFR 20.2206. Since there are no geologic repositories for high-level waste currently licensed, only six categories will be considered in this report. In 1995, the annual collective dose per reactor for light water reactor licensees (LWRs) was 199 person-cSv (person-rem). This is the same value that was reported for 1994. The annual collective dose per reactor for boiling water reactors (BWRs) was 256 person-cSv (person-rem) and, for pressurized water reactors (PWRs), it was 170 person-cSv (person-rem). Analyses of transient worker data indicate that 17,153 individuals completed work assignments at two or more licensees during the monitoring year. The dose distributions are adjusted each year to account for the duplicate reporting of transient workers by multiple licensees. In 1995, the average measurable dose calculated from reported data was 0.26 cSv (rem). The corrected dose distribution resulted in an average measurable dose of 0.32 cSv (rem).

  20. Developing Fully Coupled Dynamical Reactor Core Isolation System Models in RELAP-7 for Extended Station Black-Out Analysis

    SciTech Connect

    Haihua Zhao; Ling Zou; Hongbin Zhang; David Andrs; Richard Martineau

    2014-04-01

    The reactor core isolation cooling (RCIC) system in a boiling water reactor (BWR) provides makeup water to the reactor vessel for core cooling when the main steam lines are isolated and the normal supply of water to the reactor vessel is lost. It was one of the very few safety systems still available during the Fukushima Daiichi accidents after the tsunamis hit the plants and the system successfully delayed the core meltdown for a few days for unit 2 & 3. Therefore, detailed models for RCIC system components are indispensable to understand extended station black-out accidents (SBO) for BWRs. As part of the effort to develop the new generation reactor system safety analysis code RELAP-7, major components to simulate the RCIC system have been developed. This paper describes the models for those components such as turbine, pump, and wet well. Selected individual component test simulations and a simplified SBO simulation up to but before core damage is presented. The successful implementation of the simplified RCIC and wet well models paves the way to further improve the models for safety analysis by including more detailed physical processes in the near future.

  1. Application of dosimetry measurements to analyze the neutron activation of a stainless steel sample in a training nuclear reactor

    NASA Astrophysics Data System (ADS)

    Ródenas, J.; Gallardo, S.; Weirich, F.; Hansen, W.

    2014-11-01

    All materials present in the core of a nuclear reactor are activated by neutron irradiation. The activity so generated produces a dose around the material. This dose is a potential risk for workers in the surrounding area when materials are withdrawn from the reactor. Therefore, it is necessary to assess the activity generated and the dose produced. In previous works, neutron activation of control rods and doses around the storage pool where they are placed have been calculated for a Boiling Water Reactor using the MCNP5 code based on the Monte Carlo method. Most of the activation is produced indeed in stainless steel components of the nuclear reactor core not only control rods. In this work, a stainless steel sample is irradiated in the Training Reactor AKR-2 of the Technical University Dresden. Dose measurements around the sample have been performed for different times after the irradiation. Experimental dosimetric values are compared with results of Monte Carlo simulation of the irradiation. Comparison shows a good agreement. Hence, the activation Monte Carlo model can be considered as validated.

  2. Research reactors - an overview

    SciTech Connect

    West, C.D.

    1997-03-01

    A broad overview of different types of research and type reactors is provided in this paper. Reactor designs and operating conditions are briefly described for four reactors. The reactor types described include swimming pool reactors, the High Flux Isotope Reactor, the Mark I TRIGA reactor, and the Advanced Neutron Source reactor. Emphasis in the descriptions is placed on safety-related features of the reactors. 7 refs., 7 figs., 2 tabs.

  3. 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.

  4. NEUTRONIC REACTOR

    DOEpatents

    Ohlinger, L.A.; Wigner, E.P.; Weinberg, A.M.; Young, G.J.

    1958-09-01

    This patent relates to neutronic reactors of the heterogeneous water cooled type, and in particular to a fuel element charging and discharging means therefor. In the embodiment illustrated the reactor contains horizontal, parallel coolant tubes in which the fuel elements are disposed. A loading cart containing a magnzine for holding a plurality of fuel elements operates along the face of the reactor at the inlet ends of the coolant tubes. The loading cart is equipped with a ram device for feeding fuel elements from the magazine through the inlot ends of the coolant tubes. Operating along the face adjacent the discharge ends of the tubes there is provided another cart means adapted to receive irradiated fuel elements as they are forced out of the discharge ends of the coolant tubes by the incoming new fuel elements. This cart is equipped with a tank coataining a coolant, such as water, into which the fuel elements fall, and a hydraulically operated plunger to hold the end of the fuel element being discharged. This inveation provides an apparatus whereby the fuel elements may be loaded into the reactor, irradiated therein, and unloaded from the reactor without stopping the fiow of the coolant and without danger to the operating personnel.

  5. NUCLEAR REACTOR

    DOEpatents

    Christy, R.F.

    1958-07-15

    A nuclear reactor of the homogeneous liquid fuel type is described wherein the fissionable isotope is suspended or dissolved in a liquid moderator such as water. The reactor core is comprised essentially of a spherical vessel for containing the reactive composition surrounded by a reflector, preferably of beryllium oxide. The reactive composition may be an ordinary water solution of a soluble salt of uranium, the quantity of fissionable isotope in solution being sufficient to provide a critical mass in the vessel. The liquid fuel is stored in a tank of non-crtttcal geometry below the reactor vessel and outside of the reflector and is passed from the tank to the vessel through a pipe connecting the two by air pressure means. Neutron absorbing control and safety rods are operated within slots in the reflector adjacent to the vessel.

  6. NEUTRONIC REACTOR

    DOEpatents

    Metcalf, H.E.

    1957-10-01

    A reactor of the type which preferably uses plutonium as the fuel and a liquid moderator, preferably ordinary water, and which produces steam within the reactor core due to the heat of the chain reaction is described. In the reactor shown the fuel elements are essentially in the form of trays and are ventically stacked in spaced relationship. The water moderator is continuously supplied to the trays to maintain a constant level on the upper surfaces of the fuel element as it is continually evaporated by the heat. The steam passes out through the spaces between the fuel elements and is drawn off at the top of the core. The fuel elements are clad in aluminum to prevent deterioration thereof with consequent contamimation of the water.

  7. REACTOR CONTROL

    DOEpatents

    Fortescue, P.; Nicoll, D.

    1962-04-24

    A control system employed with a high pressure gas cooled reactor in which a control rod is positioned for upward and downward movement into the neutron field from a position beneath the reactor is described. The control rod is positioned by a coupled piston cylinder releasably coupled to a power drive means and the pressurized coolant is directed against the lower side of the piston. The coolant pressure is offset by a higher fiuid pressure applied to the upper surface of the piston and means are provided for releasing the higher pressure on the upper side of the piston so that the pressure of the coolant drives the piston upwardly, forcing the coupled control rod into the ncutron field of the reactor. (AEC)

  8. NEUTRONIC REACTORS

    DOEpatents

    Wigner, E.P.; Young, G.J.

    1958-10-14

    A method is presented for loading and unloading rod type fuel elements of a neutronic reactor of the heterogeneous, solld moderator, liquid cooled type. In the embodiment illustrated, the fuel rods are disposed in vertical coolant channels in the reactor core. The fuel rods are loaded and unloaded through the upper openings of the channels which are immersed in the coolant liquid, such as water. Unloading is accomplished by means of a coffer dam assembly having an outer sleeve which is placed in sealing relation around the upper opening. A radiation shield sleeve is disposed in and reciprocable through the coffer dam sleeve. A fuel rod engaging member operates through the axial bore in the radiation shield sleeve to withdraw the fuel rod from its position in the reactor coolant channel into the shield, the shield snd rod then being removed. Loading is accomplished in the reverse procedure.

  9. Bioconversion reactor

    DOEpatents

    McCarty, Perry L.; Bachmann, Andre

    1992-01-01

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

  10. Catalytic reactor

    SciTech Connect

    Aaron, Timothy Mark; Shah, Minish Mahendra; Jibb, Richard John

    2009-03-10

    A catalytic reactor is provided with one or more reaction zones each formed of set(s) of reaction tubes containing a catalyst to promote chemical reaction within a feed stream. The reaction tubes are of helical configuration and are arranged in a substantially coaxial relationship to form a coil-like structure. Heat exchangers and steam generators can be formed by similar tube arrangements. In such manner, the reaction zone(s) and hence, the reactor is compact and the pressure drop through components is minimized. The resultant compact form has improved heat transfer characteristics and is far easier to thermally insulate than prior art compact reactor designs. Various chemical reactions are contemplated within such coil-like structures such that as steam methane reforming followed by water-gas shift. The coil-like structures can be housed within annular chambers of a cylindrical housing that also provide flow paths for various heat exchange fluids to heat and cool components.

  11. POWER REACTOR

    DOEpatents

    Zinn, W.H.

    1958-07-01

    A fast nuclear reactor system ls described for producing power and radioactive isotopes. The reactor core is of the heterogeneous, fluid sealed type comprised of vertically arranged elongated tubular fuel elements having vertical coolant passages. The active portion is surrounded by a neutron reflector and a shield. The system includes pumps and heat exchangers for the primary and secondary coolant circuits. The core, primary coolant pump and primary heat exchanger are disposed within an irapenforate tank which is filled with the primary coolant, in this case a liquid metal such as Na or NaK, to completely submerge these elements. The tank is completely surrounded by a thick walled concrete shield. This reactor system utilizes enriched uranium or plutonium as the fissionable material, uranium or thorium as a diluent and thorium or uranium containing less than 0 7% of the U/sup 235/ isotope as a fertile material.

  12. Storage of LWR (light-water-reactor) spent fuel in air

    SciTech Connect

    Thomas, L.E.; Charlot, L.A.; Coleman, J.E. ); Knoll, R.W. )

    1989-12-01

    An experimental program is being conducted at Pacific Northwest Laboratory (PNL) to determine the oxidation response of light-water-reactor (LWR) spent fuels under conditions appropriate to fuel storage in air. The program is designed to investigate several independent variables that might affect the oxidation behavior of spent fuel. Included are temperature (135 to 230{degree}C), fuel burnup (to about 34 MWd/kgM), reactor type (pressurized and boiling water reactors), moisture level in the air, and the presence of a high gamma field. In continuing tests with declad spent fuel and nonirradiated UO{sub 2} specimens, oxidation rates were monitored by weight-gain measurements and the microstructures of subsamples taken during the weighing intervals were characterized by several analytical methods. The oxidation behavior indicated by weight gain and time to form powder will be reported in Volume III of this series. The characterization results obtained from x-ray diffractometry, transmission electron microscopy, scanning electron microscopy, and Auger electron spectrometry of oxidized fuel samples are presented in this report. 28 refs., 21 figs., 3 tabs.

  13. Environmentally assisted cracking in Light Water Reactors: Semiannual report, October 1994--March 1995. Volume 20

    SciTech Connect

    Chung, H.M.; Chopra, O.K.; Gavenda, D.J.; Hins, A.G.; Kassner, T.F.; Ruther, W.E.; Shack, W.J.; Soppet, W.K.

    1996-01-01

    This report summarizes work performed by Argonne National Laboratory on fatigue and environmentally assisted cracking (EAC) in light water reactors (LWRS) from October 1994 to March 1995. Topics that have been investigated include (a) fatigue of carbon and low-alloy steel used in reactor piping and pressure vessels, (b) EAC of Alloy 600 and 690, and (c) irradiation-assisted stress corrosion cracking (IASCC) of Type 304 SS. Fatigue tests were conducted on ferritic steels in water with several dissolvedoxygen (DO) concentrations to determine whether a slow strain rate applied during different portions of a tensile-loading cycle are equally effective in decreasing fatigue life. Tensile properties and microstructures of several heats of Alloy 600 and 690 were characterized for correlation with EAC of the alloys in simulated LWR environments. Effects of DO and electrochemical potential on susceptibility to intergranular cracking of high- and commercial-purity Type 304 SS specimens from control-blade absorber tubes and a control-blade sheath irradiated in boiling water reactors were determined in slow-strain-rate-tensile tests at 289{degrees}C. 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.

  14. NEUTRONIC REACTOR

    DOEpatents

    Fermi, E.; Szilard, L.

    1957-09-24

    Reactors of the type employing plates of natural uranium in a moderator are discussed wherein the plates are um-formly disposed in parallel relationship to each other thereby separating the moderator material into distinct and individual layers. Each plate has an uninterrupted sunface area substantially equal to the cross-sectional area of the active portion of the reactor, the particular size of the plates and the volume ratio of moderator to uranium required to sustain a chain reaction being determinable from the known purity of these materials and other characteristics such as the predictable neutron losses due to the formation of radioactive elements of extremely high neutron capture cross section.

  15. NEUTRONIC REACTOR

    DOEpatents

    Wigner, E.P.; Weinberg, A.W.; Young, G.J.

    1958-04-15

    A nuclear reactor which uses uranium in the form of elongated tubes as fuel elements and liquid as a coolant is described. Elongated tubular uranium bodies are vertically disposed in an efficient neutron slowing agent, such as graphite, for example, to form a lattice structure which is disposed between upper and lower coolant tanks. Fluid coolant tubes extend through the uranium bodies and communicate with the upper and lower tanks and serve to convey the coolant through the uranium body. The reactor is also provided with means for circulating the cooling fluid through the coolant tanks and coolant tubes, suitable neutron and gnmma ray shields, and control means.

  16. Etude du Photochromisme et de la Photorefractivite dans le Poly

    NASA Astrophysics Data System (ADS)

    Ghailane, Fatima

    1995-11-01

    Nous avons etudie la possibilite d'utiliser un materiau organique, le poly(vinylcarbazole) comme milieu de stockage optique de masse en temps reel. Ce materiau dope aux photochromes presente un potentiel non negligeable pour l'holographie de volume en temps reel. Plusiers cycles, Ecriture-Lecture-Effacement, ont ete enregistres dans cette matrice polymerique avec une resolution assez elevee. Une etude a ete menee afin d'obtenir la valeur des parametres comme l'epaisseur du film et l'intensite d'ecriture pour un angle d'enregistrement theta_{ acute ecriture} donne qui permettent d'obtenir les meilleures efficacites diffractionelles tout en preservant ce milieu de tout effet de fatigue ou de degradation. Ce polymere devient photorefractif une fois melange avec un bon accepteur de charge, le trinitrofluorenone et un bon chromophore non-lineaire, le disperse orange 25. Nous avons etudie experimentalement la conductivite en obscurite et la photoconductivite de ce materiau. Des simulations numeriques ont ete realisees en se basant sur le modele de Poole-Frenkel. Nous avons aussi etudie l'implantation de faisceaux d'ions d'oxygene d'energie 200 keV pour inscrire dans le PVCz un guide d'onde plan a profil d'indice. Une etude ESCA complete cette partie afin de savoir les modifications induites par implementation ionique a l'interieur du film polymerique.

  17. 75 FR 80854 - Toshiba Corporation; Acceptance for Docketing of an Application for Renewal of the U.S. Advanced...

    Federal Register 2010, 2011, 2012, 2013, 2014

    2010-12-23

    ... Boiling Water Reactor Design Certification On November 2, 2010, Toshiba Corporation (Toshiba) submitted an....S. Advanced Boiling Water Reactor (ABWR) in accordance with the requirements contained in 10...

  18. Neutronic reactor

    DOEpatents

    Carleton, John T.

    1977-01-25

    A graphite-moderated nuclear reactor includes channels between blocks of graphite and also includes spacer blocks between adjacent channeled blocks with an axis of extension normal to that of the axis of elongation of the channeled blocks to minimize changes in the physical properties of the graphite as a result of prolonged neutron bombardment.

  19. Sonochemical Reactors.

    PubMed

    Gogate, Parag R; Patil, Pankaj N

    2016-10-01

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

  20. NEUTRONIC REACTORS

    DOEpatents

    Anderson, H.L.

    1958-10-01

    The design of control rods for nuclear reactors are described. In this design the control rod consists essentially of an elongated member constructed in part of a neutron absorbing material and having tube means extending therethrough for conducting a liquid to cool the rod when in use.

  1. NEUTRONIC REACTOR

    DOEpatents

    Creutz, E.C.; Ohlinger, L.A.; Weinberg, A.M.; Wigner, E.P.; Young, G.J.

    1959-10-27

    BS>A reactor cooled by water, biphenyl, helium, or other fluid with provision made for replacing the fuel rods with the highest plutonium and fission product content without disassembling the entire core and for promptly cooling the rods after their replacement in order to prevent build-up of heat from fission product activity is described.

  2. Neutronic reactor

    DOEpatents

    Wende, Charles W. J.

    1976-08-17

    A safety rod for a nuclear reactor has an inner end portion having a gamma absorption coefficient and neutron capture cross section approximately equal to those of the adjacent shield, a central portion containing materials of high neutron capture cross section and an outer end portion having a gamma absorption coefficient at least equal to that of the adjacent shield.

  3. Light Water Reactor Sustainability Program BWR High-Fluence Material Project: Assessment of the Role of High-Fluence on the Efficiency of HWC Mitigation on SCC Crack Growth Rates

    SciTech Connect

    Sebastien Teysseyre

    2014-04-01

    As nuclear power plants age, the increasing neutron fluence experienced by stainless steels components affects the materials resistance to stress corrosion cracking and fracture toughness. The purpose of this report is to identify any new issues that are expected to rise as boiling water reactor power plants reach the end of their initial life and to propose a path forward to study such issues. It has been identified that the efficiency of hydrogen water chemistry mitigation technology may decrease as fluence increases for high-stress intensity factors. This report summarizes the data available to support this hypothesis and describes a program plan to determine the efficiency of hydrogen water chemistry as a function of the stress intensity factor applied and fluence. This program plan includes acquisition of irradiated materials, generation of material via irradiation in a test reactor, and description of the test plan. This plan offers three approaches, each with an estimated timetable and budget.

  4. NUCLEAR REACTOR

    DOEpatents

    Grebe, J.J.

    1959-12-15

    A reactor which is particularly adapted tu serve as a heat source for a nuclear powered alrcraft or rocket is described. The core of this reactor consists of a porous refractory modera;or body which is impregnated with fissionable nuclei. The core is designed so that its surface forms tapered inlet and outlet ducts which are separated by the porous moderator body. In operation a gaseous working fluid is circulated through the inlet ducts to the surface of the moderator, enters and passes through the porous body, and is heated therein. The hot gas emerges into the outlet ducts and is available to provide thrust. The principle advantage is that tremendous quantities of gas can be quickly heated without suffering an excessive pressure drop.

  5. REACTOR UNLOADING

    DOEpatents

    Leverett, M.C.

    1958-02-18

    This patent is related to gas cooled reactors wherein the fuel elements are disposed in vertical channels extending through the reactor core, the cooling gas passing through the channels from the bottom to the top of the core. The invention is a means for unloading the fuel elements from the core and comprises dump values in the form of flat cars mounted on wheels at the bottom of the core structure which support vertical stacks of fuel elements. When the flat cars are moved, either manually or automatically, for normal unloading purposes, or due to a rapid rise in the reproduction ratio within the core, the fuel elements are permtted to fall by gravity out of the core structure thereby reducing the reproduction ratio or stopping the reaction as desired.

  6. NEUTRONIC REACTOR

    DOEpatents

    Wade, E.J.

    1958-09-16

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

  7. REACTOR MONITORING

    DOEpatents

    Bugbee, S.J.; Hanson, V.F.; Babcock, D.F.

    1959-02-01

    A neutron density inonitoring means for reactors is described. According to this invention a tunnel is provided beneath and spaced from the active portion of the reactor and extends beyond the opposite faces of the activc portion. Neutron beam holes are provided between the active portion and the tunnel and open into the tunnel near the middle thereof. A carriage operates back and forth in the tunnel and is adapted to convey a neutron detector, such as an ion chamber, and position it beneath one of the neutron beam holes. This arrangement affords convenient access of neutron density measuring instruments to a location wherein direct measurement of neutron density within the piles can be made and at the same time affords ample protection to operating personnel.

  8. Nuclear reactor

    DOEpatents

    Wade, Elman E.

    1979-01-01

    A nuclear reactor including two rotatable plugs and a positive top core holddown structure. The top core holddown structure is divided into two parts: a small core cover, and a large core cover. The small core cover, and the upper internals associated therewith, are attached to the small rotating plug, and the large core cover, with its associated upper internals, is attached to the large rotating plug. By so splitting the core holddown structures, under-the-plug refueling is accomplished without the necessity of enlarging the reactor pressure vessel to provide a storage space for the core holddown structure during refueling. Additionally, the small and large rotating plugs, and their associated core covers, are arranged such that the separation of the two core covers to permit rotation is accomplished without the installation of complex lifting mechanisms.

  9. 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.

  10. Neutronic reactor

    DOEpatents

    Lewis, Warren R.

    1978-05-30

    A graphite-moderated, water-cooled nuclear reactor including a plurality of rectangular graphite blocks stacked in abutting relationship in layers, alternate layers having axes which are normal to one another, alternate rows of blocks in alternate layers being provided with a channel extending through the blocks, said channeled blocks being provided with concave sides and having smaller vertical dimensions than adjacent blocks in the same layer, there being nuclear fuel in the channels.

  11. NUCLEAR REACTORS

    DOEpatents

    Long, E.; Ashby, J.W.

    1958-09-16

    ABS>A graphite moderator structure is presented for a nuclear reactor compriscd of an assembly of similarly orientated prismatic graphite blocks arranged on spaced longitudinal axes lying in common planes wherein the planes of the walls of the blocks are positioned so as to be twisted reintive to the planes of said axes so thatthe unlmpeded dtrect paths in direction wholly across the walls of the blocks are limited to the width of the blocks plus spacing between the blocks.

  12. NUCLEAR REACTOR

    DOEpatents

    Grebe, J.J.

    1961-01-24

    A core structure for neutronic reactors adapted for the propulsion of aircraft and rockets is offered. The core is designed for cooling by gaseous media, and comprises a plurality of hollow tapered tubular segments of a porous moderating material impregniated with fissionable fuel nested about a common axis. Alternate ends of the segments are joined. In operation a coolant gas passes through the porous structure and is heated.

  13. NUCLEAR REACTORS

    DOEpatents

    Koch, L.J.; Rice, R.E. Jr.; Denst, A.A.; Rogers, A.J.; Novick, M.

    1961-12-01

    An active portion assembly for a fast neutron reactor is described wherein physical distortions resulting in adverse changes in the volume-to-mass ratio are minimized. A radially expandable locking device is disposed within a cylindrical tube within each fuel subassembly within the active portion assembly, and clamping devices expandable toward the center of the active portion assembly are disposed around the periphery thereof. (AEC)

  14. REACTOR CONTROL

    DOEpatents

    Ruano, W.J.

    1957-12-10

    This patent relates to nuclear reactors of the type which utilize elongited rod type fuel elements immersed in a liquid moderator and shows a design whereby control of the chain reaction is obtained by varying the amount of moderator or reflector material. A central tank for containing liquid moderator and fuel elements immersed therein is disposed within a surrounding outer tank providing an annular space between the two tanks. This annular space is filled with liquid moderator which functions as a reflector to reflect neutrons back into the central reactor tank to increase the reproduction ratio. Means are provided for circulating and cooling the moderator material in both tanks and additional means are provided for controlling separately the volume of moderator in each tank, which latter means may be operated automatically by a neutron density monitoring device. The patent also shows an arrangement for controlling the chain reaction by injecting and varying an amount of poisoning material in the moderator used in the reflector portion of the reactor.

  15. Space reactors

    NASA Astrophysics Data System (ADS)

    Ranken, W. A.

    1983-01-01

    Progress in design studies and technology for the SP-100 Project - successor to the Space Power Advanced Reactor (SPAR) Project - is reported for the period October 1, 1981 to March 31, 1982. The basis for selecting a high-temperature, UO2-fueled, heat-pipe-cooled reactor with a thermoelectric conversion system as the 100/kW-sub e/ reference design has been reviewed. Although no change has been made in the general concept, design studies have been done to investigate various reactor/conversion system coupling methods and core design modifications. Thermal and mechanical finite element modeling and three dimensional Monte Carlo analysis of a core with individual finned fuel elements are reported. Studies of unrestrained fuel irradiation data are discussed that are relevant both to the core modeling work and to the design and fabrication of the first in-pile irradiation test, which is also reported. Work on lithium-filled core heat pipe development is described, including the attainment of 15.6 kW/sub t/ operation at 1525 K for a 2-m-long heat pipe with a 15.7-mm outside diameter. The successful operation of a 5.5-m-long, lightweight potassium/titanium heat pipe at 760 K is described, and test results of a thermoelectric module with GaP-modified SiGe thermoelectric elements are presented.

  16. Influence of temperature histories during reactor startup periods on microstructural evolution and mechanical properties of austenitic stainless steel irradiated with neutrons

    NASA Astrophysics Data System (ADS)

    Kasahara, Shigeki; Kitsunai, Yuji; Chimi, Yasuhiro; Chatani, Kazuhiro; Koshiishi, Masato; Nishiyama, Yutaka

    2016-11-01

    This paper addresses influence of two different temperature profiles during startup periods in the Japan Materials Testing Reactor and a boiling water reactor upon microstructural evolution and mechanical properties of austenitic stainless steel irradiated with neutrons to about 1 dpa and 3 dpa. One of the temperature profiles was that the specimens experienced neutron irradiation in both reactors, under which the irradiation temperature transiently increased to 290 °C from room temperature with increasing reactor power during reactor startup periods. Another was that the specimens were pre-heated to about 150 °C prior to the irradiation to suppress the transient temperature increase. Tensile tests at 290 °C and Vickers hardness tests at room temperature were carried out, and their microstructures were observed by FEG-TEM. Difference of the temperature profiles was observed obviously in interstitial cluster formation, in particular, growth of Frank loops. Although influence of neutron irradiation involving transient temperature increase to 290 °C from room temperature on the yield strength and the Vickers hardness is buried in the trend curves of existing data, the influence was also found certainly in increment of in yield strength, existence of modest yield drop, and loss of strain hardening capacity and ductility. As a result, Frank loops, which were observed in austenitic stainless steel irradiated at doses of 1 dpa or more, seemed to have important implications regarding the interpretation of not irradiation hardening, but deformation of the austenitic stainless steel.

  17. An Assessment of Remote Visual Methods to Detect Cracking in Reactor Components

    SciTech Connect

    Cumblidge, Stephen E.; Anderson, Michael T.; Doctor, Steven R.; Simonen, Fredric A.; Elliot, Anthony J.

    2008-01-01

    Recently, the U.S. nuclear industry has proposed replacing current volumetric and/or surface examinations of certain components in commercial nuclear power plants, as required by the American Society of Mechanical Engineers Boiler and Pressure Vessel Code Section XI, “Inservice Inspection of Nuclear Power Plant Components,” with a simpler visual testing (VT) method. The advantages of VT are that these tests generally involve much less radiation exposure and time to perform the examination than do volumetric examinations such as ultrasonic testing. The issues relative to the reliability of VT in determining the structural integrity of reactor components were examined. Some piping and pressure vessel components in a nuclear power station are examined using VT as they are either in high radiation fields or component geometry precludes the use of ultrasonic testing (UT) methodology. Remote VT with radiation-hardened video systems has been used by nuclear utilities to find cracks in pressure vessel cladding in pressurized water reactors, core shrouds in boiling water reactors, and to investigate leaks in piping and reactor components. These visual tests are performed using a wide variety of procedures and equipment. The techniques for remote VT use submersible closed-circuit video cameras to examine reactor components and welds. PNNL conducted a parametric study that examined the important variables influencing the effectiveness of a remote visual test. Tested variables included lighting techniques, camera resolution, camera movement, and magnification. PNNL also conducted a limited laboratory test using a commercial visual testing camera system to experimentally determine the ability of the camera system to detect cracks of various widths under ideal conditions. The results of these studies and their implications are presented in this paper.

  18. 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.

  19. Occupational radiation exposure at commercial nuclear power reactors and other facilities 1994. Twenty-seventh annual report

    SciTech Connect

    Thomas, M.L.; Hagemeyer, D.

    1996-01-01

    This report summarizes the occupational exposure data that are maintained in the U.S. Nuclear Regulatory Commission`s (NRC) Radiation Exposure Information and Reporting System (REIRS). Annual reports for 1994 were received from a total of 303 NRC licensees, of which 109 were operators of nuclear power reactors in commercial operation. Compilations of the reports submitted by the 303 licensees indicated that 152,028 individuals were monitored, 79,780 of whom received a measurable dose. The collective dose incurred by these individuals was 24,740 person-cSv (person-rem){sup 2} which represents a 15% decrease from the 1993 value. The number of workers receiving a measurable dose also decreased, resulting in the average measurable dose of 0.31 cSv (rem) for 1994. The average measurable dose is defined to be the total collective dose (TEDE) divided by the number of workers receiving a measurable dose. These figures have been adjusted to account for transient reactor workers. In 1994, the annual collective dose per reactor for light water reactor licensees (LWRs) was 198 person-cSv (person-rem). This represents a 18% decrease from the 1993 value of 242 person-cSv (person-rem). The annual collective dose per reactor for boiling water reactors (BWRs) was 327 person-cSv (person-rem) and, for pressurized water reactors (PWRs), it was 131 person-cSv (person-rem). Analyses of transient worker data indicate that 18,178 individuals completed work assignments at two or more licensees during the monitoring year. The dose distributions are adjusted each year to account for the duplicate reporting of transient workers by multiple licensees. In 1994, the average measurable dose calculated from reported data was 0.28 cSv (rem). The corrected dose distribution resulted in an average measurable dose of 0.31 cSv (rem).

  20. Prediction of the Isotopic Composition of UO(Sub 2) Fuel from a BWR: Analysis of the DU1 Sample from the Dodeward Reactor

    SciTech Connect

    Murphy, B.D.

    1998-01-01

    As part of a larger program to study mixed-oxide fuel subject to high burnup, some UO{sub 2} samples were exposed and analyzed. This report discusses results from the analysis of a UO{sub 2} sample that was burned in a boiling-water reactor (BWR) to approximately 57 GWd/t. The sample enrichment was high (a U{sup 235} content of 4.94%) relative to the surrounding UO{sub 2} fuel. The isotopic content of the discharged sample was determined experimentally (both actinides and fission products), and the measured concentrations are compared with calculated values using both the Oak Ridge National Laboratory SCALE system and the HELIOS code system that is marketed by Scandpower. Because the sample enrichment differed from that of the surrounding fuel, this test was a rather stringent test of the simulation models. These results are discussed, as are the general issues surrounding the simulation of fuel burnup in a BWR.

  1. Nuclear Reactors. Revised.

    ERIC Educational Resources Information Center

    Hogerton, John F.

    This publication is one of a series of information booklets for the general public published by the United States Atomic Energy Commission. Among the topics discussed are: How Reactors Work; Reactor Design; Research, Teaching, and Materials Testing; Reactors (Research, Teaching and Materials); Production Reactors; Reactors for Electric Power…

  2. Modeling of a Flooding Induced Station Blackout for a Pressurized Water Reactor Using the RISMC Toolkit

    SciTech Connect

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

    2011-07-01

    In the Risk Informed Safety Margin Characterization (RISMC) approach we want to understand not just the frequency of an event like core damage, but how close we are (or are not) to key safety-related events and how might we increase our safety margins. The RISMC Pathway uses the probabilistic margin approach to quantify impacts to reliability and safety by coupling both probabilistic (via stochastic simulation) and mechanistic (via physics models) approaches. This coupling takes place through the interchange of physical parameters and operational or accident scenarios. In this paper we apply the RISMC approach to evaluate the impact of a power uprate on a pressurized water reactor (PWR) for a tsunami-induced flooding test case. This analysis is performed using the RISMC toolkit: RELAP-7 and RAVEN codes. RELAP-7 is the new generation of system analysis codes that is responsible for simulating the thermal-hydraulic dynamics of PWR and boiling water reactor systems. RAVEN has two capabilities: to act as a controller of the RELAP-7 simulation (e.g., system activation) and to perform statistical analyses (e.g., run multiple RELAP-7 simulations where sequencing/timing of events have been changed according to a set of stochastic distributions). By using the RISMC toolkit, we can evaluate how power uprate affects the system recovery measures needed to avoid core damage after the PWR lost all available AC power by a tsunami induced flooding. The simulation of the actual flooding is performed by using a smooth particle hydrodynamics code: NEUTRINO.

  3. Nuclear reactor

    DOEpatents

    Pennell, William E.; Rowan, William J.

    1977-01-01

    A nuclear reactor in which the core components, including fuel-rod assemblies, control-rod assemblies, fertile rod-assemblies, and removable shielding assemblies, are supported by a plurality of separate inlet modular units. These units are referred to as inlet module units to distinguish them from the modules of the upper internals of the reactor. The modular units are supported, each removable independently of the others, in liners in the supporting structure for the lower internals of the reactor. The core assemblies are removably supported in integral receptacles or sockets of the modular units. The liners, units, sockets and assmblies have inlet openings for entry of the fluid. The modular units are each removably mounted in the liners with fluid seals interposed between the opening in the liner and inlet module into which the fluid enters and the upper and lower portion of the liner. Each assembly is similarly mounted in a corresponding receptacle with fluid seals interposed between the openings where the fluid enters and the lower portion of the receptacle or fitting closely in these regions. As fluid flows along each core assembly a pressure drop is produced along the fluid so that the fluid which emerges from each core assembly is at a lower pressure than the fluid which enters the core assembly. However because of the seals interposed in the mountings of the units and assemblies the pressures above and below the units and assemblies are balanced and the units are held in the liners and the assemblies are held in the receptacles by their weights as they have a higher specific gravity than the fluid. The low-pressure spaces between each module and its liner and between each core assembly and its module is vented to the low-pressure regions of the vessel to assure that fluid which leaks through the seals does not accumulate and destroy the hydraulic balance.

  4. NUCLEAR REACTORS

    DOEpatents

    Long, E.; Ashley, J.W.

    1958-12-16

    A graphite moderator structure is described for a gas-cooled nuclear reactor having a vertical orlentation wherein the structure is physically stable with regard to dlmensional changes due to Wigner growth properties of the graphite, and leakage of coolant gas along spaces in the structure is reduced. The structure is comprised of stacks of unlform right prismatic graphite blocks positioned in layers extending in the direction of the lengths of the blocks, the adjacent end faces of the blocks being separated by pairs of tiles. The blocks and tiles have central bores which are in alignment when assembled and are provided with cooperatlng keys and keyways for physical stability.

  5. ELECTRONUCLEAR REACTOR

    DOEpatents

    Lawrence, E.O.; McMillan, E.M.; Alvarez, L.W.

    1960-04-19

    An electronuclear reactor is described in which a very high-energy particle accelerator is employed with appropriate target structure to produce an artificially produced material in commercial quantities by nuclear transformations. The principal novelty resides in the combination of an accelerator with a target for converting the accelerator beam to copious quantities of low-energy neutrons for absorption in a lattice of fertile material and moderator. The fertile material of the lattice is converted by neutron absorption reactions to an artificially produced material, e.g., plutonium, where depleted uranium is utilized as the fertile material.

  6. REACTOR COMPONETN

    DOEpatents

    Creutz, E.C.

    1959-10-27

    A reactor fuel element comprised of a slug of fissionable material disposed in a sheath of corrosion resistantmaterial is described. The sheath is in the form of a tubular container closed at one end and is in tight-fitting engagement with the peripheral sunface of the slug. An inner cap is insented into the open end of the sheath against the slug, which end is then bent around the inner cap and welded thereto. An outer cap is then welded around its peripheny to the bent portion of the container.

  7. Photocatalytic reactor

    DOEpatents

    Bischoff, B.L.; Fain, D.E.; Stockdale, J.A.D.

    1999-01-19

    A photocatalytic reactor is described for processing selected reactants from a fluid medium comprising at least one permeable photocatalytic membrane having a photocatalytic material. The material forms an area of chemically active sites when illuminated by light at selected wavelengths. When the fluid medium is passed through the illuminated membrane, the reactants are processed at these sites separating the processed fluid from the unprocessed fluid. A light source is provided and a light transmitting means, including an optical fiber, for transmitting light from the light source to the membrane. 4 figs.

  8. Dynamic Comparison of Three- and Four-Equation Reactor Core Models in a Full-Scope Power Plant Training Simulator

    SciTech Connect

    Espinosa-Paredes, Gilberto; Alvarez-Ramirez, Jose; Nunez-Carrera, Alejandro; Garcia-Gutierrez, Alfonso; Martinez-Mendez, Elizabeth Jeannette

    2004-02-15

    A comparative analysis of the dynamic behavior of a boiling water reactor in a full-scope power plant simulator for operator training is presented. Three- and four-equation reactor core models were used to examine three transients following tests described in acceptance test procedures: scram, loss of feedwater flow, and closure of main isolation valves. The three-equation model consists of water and steam mixture momentum, including mass and energy balances. The four-equation model is based on liquid and gas phase mass balances, together with a drift-flux approach for the analysis of phase separation. Analysis of the models showed that the scram transient was slightly different for three- and four-equation models. The drift-flux effects can explain such differences. Regarding the loss-of-feedwater transient, the predicted steam flow after scram is larger for the three-equation model. Finally, for the transient related to the closure of main steam isolation valves, the three-equation model provides slightly different results for the pressure change, which affects reactor level behavior.

  9. Hybrid adsorptive membrane reactor

    NASA Technical Reports Server (NTRS)

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

    2011-01-01

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

  10. Hybrid adsorptive membrane reactor

    DOEpatents

    Tsotsis, Theodore T [Huntington Beach, CA; Sahimi, Muhammad [Altadena, CA; Fayyaz-Najafi, Babak [Richmond, CA; Harale, Aadesh [Los Angeles, CA; Park, Byoung-Gi [Yeosu, KR; Liu, Paul K. T. [Lafayette Hill, PA

    2011-03-01

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

  11. Control Means for Reactor

    DOEpatents

    Manley, J. H.

    1961-06-27

    An apparatus for controlling a nuclear reactor includes a tank just below the reactor, tubes extending from the tank into the reactor, and a thermally expansible liquid neutron absorbent material in the tank. The liquid in the tank is exposed to a beam of neutrons from the reactor which heats the liquid causing it to expand into the reactor when the neutron flux in the reactor rises above a predetermincd danger point. Boron triamine may be used for this purpose.

  12. Reactor Dosimetry State of the Art 2008

    NASA Astrophysics Data System (ADS)

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

    2009-08-01

    Oral session 1: Retrospective dosimetry. Retrospective dosimetry of VVER 440 reactor pressure vessel at the 3rd unit of Dukovany NPP / M. Marek ... [et al.]. Retrospective dosimetry study at the RPV of NPP Greifswald unit 1 / J. Konheiser ... [et al.]. Test of prototype detector for retrospective neutron dosimetry of reactor internals and vessel / K. Hayashi ... [et al.]. Neutron doses to the concrete vessel and tendons of a magnox reactor using retrospective dosimetry / D. A. Allen ... [et al.]. A retrospective dosimetry feasibility study for Atucha I / J. Wagemans ... [et al.]. Retrospective reactor dosimetry with zirconium alloy samples in a PWR / L. R. Greenwood and J. P. Foster -- Oral session 2: Experimental techniques. Characterizing the Time-dependent components of reactor n/y environments / P. J. Griffin, S. M. Luker and A. J. Suo-Anttila. Measurements of the recoil-ion response of silicon carbide detectors to fast neutrons / F. H. Ruddy, J. G. Seidel and F. Franceschini. Measurement of the neutron spectrum of the HB-4 cold source at the high flux isotope reactor at Oak Ridge National Laboratory / J. L. Robertson and E. B. Iverson. Feasibility of cavity ring-down laser spectroscopy for dose rate monitoring on nuclear reactor / H. Tomita ... [et al.]. Measuring transistor damage factors in a non-stable defect environment / D. B. King ... [et al.]. Neutron-detection based monitoring of void effects in boiling water reactors / J. Loberg ... [et al.] -- Poster session 1: Power reactor surveillance, retrospective dosimetry, benchmarks and inter-comparisons, adjustment methods, experimental techniques, transport calculations. Improved diagnostics for analysis of a reactor pulse radiation environment / S. M. Luker ... [et al.]. Simulation of the response of silicon carbide fast neutron detectors / F. Franceschini, F. H. Ruddy and B. Petrović. NSV A-3: a computer code for least-squares adjustment of neutron spectra and measured dosimeter responses / J. G

  13. ADAPTATION OF CRACK GROWTH DETECTION TECHNIQUES TO US MATERIAL TEST REACTORS

    SciTech Connect

    A. Joseph Palmer; Sebastien P. Teysseyre; Kurt L. Davis; Gordon Kohse; Yakov Ostrovsky; David M. Carpenter; Joy L. Rempe

    2015-04-01

    A key component in evaluating the ability of Light Water Reactors to operate beyond 60 years is characterizing the degradation of materials exposed to radiation and various water chemistries. Of particular concern is the response of reactor materials to Irradiation Assisted Stress Corrosion Cracking (IASCC). Some test reactors outside the United States, such as the Halden Boiling Water Reactor (HBWR), have developed techniques to measure crack growth propagation during irradiation. The basic approach is to use a custom-designed compact loading mechanism to stress the specimen during irradiation, while the crack in the specimen is monitored in-situ using the Direct Current Potential Drop (DCPD) method. In 2012 the US Department of Energy commissioned the Idaho National Laboratory and the MIT Nuclear Reactor Laboratory (MIT NRL) to take the basic concepts developed at the HBWR and adapt them to a test rig capable of conducting in-pile IASCC tests in US Material Test Reactors. The first two and half years of the project consisted of designing and testing the loader mechanism, testing individual components of the in-pile rig and electronic support equipment, and autoclave testing of the rig design prior to insertion in the MIT Reactor. The load was applied to the specimen by means of a scissor like mechanism, actuated by a miniature metal bellows driven by pneumatic pressure and sized to fit within the small in-core irradiation volume. In addition to the loader design, technical challenges included developing robust connections to the specimen for the applied current and voltage measurements, appropriate ceramic insulating materials that can endure the LWR environment, dealing with the high electromagnetic noise environment of a reactor core at full power, and accommodating material property changes in the specimen, due primarily to fast neutron damage, which change the specimen resistance without additional crack growth. The project culminated with an in

  14. Environmentally assisted cracking in Light Water Reactors. Volume 16: Semiannual report, October 1992--March 1993

    SciTech Connect

    Chung, H.M.; Chopra, O.K.; Ruther, W.E.; Kassner, T.F.; Michaud, W.F.; Park, J.Y.; Sanecki, J.E.; Shack, W.J.

    1993-09-01

    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 October 1992 to March 1993. Fatigue and EAC of piping, pressure vessels, and core components in LWRs are important concerns as extended reactor lifetimes are envisaged. Topics that have been investigated include (1) fatigue of low-alloy steel used in piping, steam generators, and reactor pressure vessels. (2) EAC of cast stainless steels (SSs), (3) radiation-induced segregation and irradiation-assisted stress corrosion cracking of Type 304 SS after accumulation of relatively high fluence, and (4) EAC of low-alloy steels. 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 and chromium-nickel-plated A533-Gr B steel in simulated boiling-water reactor (BWR) water at 289{degrees}C. The data were compared with predictions based on crack growth correlations for ferritic steels in oxygenated water and 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 BWRs were studied by Auger electron spectroscopy and scanning electron microscopy.

  15. 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.

  16. Evaluation of a Method for Remote Detection of Fuel Relocation Outside the Original Core Volumes of Fukushima Reactor Units 1-3

    SciTech Connect

    Douglas W. Akers; Edwin A. Harvego

    2012-08-01

    This paper presents the results of a study to evaluate the feasibility of remotely detecting and quantifying fuel relocation from the core to the lower head, and to regions outside the reactor vessel primary containment of the Fukushima 1-3 reactors. The goals of this study were to determine measurement conditions and requirements, and to perform initial radiation transport sensitivity analyses for several potential measurement locations inside the reactor building. The radiation transport sensitivity analyses were performed based on reactor design information for boiling water reactors (BWRs) similar to the Fukushima reactors, ORIGEN2 analyses of 3-cycle BWR fuel inventories, and data on previously molten fuel characteristics from TMI- 2. A 100 kg mass of previously molten fuel material located on the lower head of the reactor vessel was chosen as a fuel interrogation sensitivity target. Two measurement locations were chosen for the transport analyses, one inside the drywell and one outside the concrete biological shield surrounding the drywell. Results of these initial radiation transport analyses indicate that the 100 kg of previously molten fuel material may be detectable at the measurement location inside the drywell, but that it is highly unlikely that any amount of fuel material inside the RPV will be detectable from a location outside the concrete biological shield surrounding the drywell. Three additional fuel relocation scenarios were also analyzed to assess detection sensitivity for varying amount of relocated material in the lower head of the reactor vessel, in the control rods perpendicular to the detector system, and on the lower head of the drywell. Results of these analyses along with an assessment of background radiation effects and a discussion of measurement issues, such as the detector/collimator design, are included in the paper.

  17. NEUTRONIC REACTOR

    DOEpatents

    Stewart, H.B.

    1958-12-23

    A nuclear reactor of the type speclfically designed for the irradiation of materials is discussed. In this design a central cyllndrical core of moderating material ls surrounded by an active portlon comprlsed of an annular tank contalning fissionable material immersed ln a liquid moderator. The active portion ls ln turn surrounded by a reflector, and a well ls provided in the center of the core to accommodate the materlals to be irradiated. The over-all dimensions of the core ln at least one plane are equal to or greater than twice the effective slowing down length and equal to or less than twlce the effective diffuslon length for neutrons in the core materials.

  18. NEUTRONIC REACTOR

    DOEpatents

    Daniels, F.

    1962-12-18

    A power plant is described comprising a turbine and employing round cylindrical fuel rods formed of BeO and UO/sub 2/ and stacks of hexagonal moderator blocks of BeO provided with passages that loosely receive the fuel rods so that coolant may flow through the passages over the fuels to remove heat. The coolant may be helium or steam and fiows through at least one more heat exchanger for producing vapor from a body of fluid separate from the coolant, which fluid is to drive the turbine for generating electricity. By this arrangement the turbine and directly associated parts are free of particles and radiations emanating from the reactor. (AEC)

  19. NEUTRONIC REACTOR

    DOEpatents

    McGarry, R.J.

    1958-04-22

    Fluid-cooled nuclear reactors of the type that utilize finned uranium fuel elements disposed in coolant channels in a moderater are described. The coolant channels are provided with removable bushings composed of a non- fissionable material. The interior walls of the bushings have a plurality of spaced, longtudinal ribs separated by grooves which receive the fins on the fuel elements. The lands between the grooves are spaced from the fuel elements to form flow passages, and the size of the now passages progressively decreases as the dlstance from the center of the core increases for the purpose of producing a greater cooling effect at the center to maintain a uniform temperature throughout the core.

  20. Nuclear reactor neutron shielding

    DOEpatents

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

    2017-09-12

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

  1. Reactor and method of operation

    DOEpatents

    Wheeler, John A.

    1976-08-10

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

  2. Nuclear reactor

    DOEpatents

    Yant, Howard W.; Stinebiser, Karl W.; Anzur, Gregory C.

    1977-01-01

    A nuclear reactor, particularly a liquid-metal breeder reactor, whose upper internals include outlet modules for channeling the liquid-metal coolant from selected areas of the outlet of the core vertically to the outlet plenum. The modules are composed of a highly-refractory, high corrosion-resistant alloy, for example, INCONEL-718. Each module is disposed to confine and channel generally vertically the coolant emitted from a subplurality of core-component assemblies. Each module has a grid with openings, each opening disposed to receive the coolant from an assembly of the subplurality. The grid in addition serves as a holdown for the assemblies of the corresponding subplurality preventing their excessive ejection upwardly from the core. In the region directly over the core the outlet modules are of such peripheral form that they nest forming a continuum over the core-component assemblies whose outlet coolant they confine. Each subassembly includes a chimney which confines the coolant emitted by its corresponding subassemblies to generally vertical flow between the outlet of the core and the outlet plenum. Each subplurality of assemblies whose emitted coolant is confined by an outlet module includes assemblies which emit lower-temperature coolant, for example, a control-rod assembly, or fertile assemblies, and assemblies which emit coolant of substantially higher temperature, for example, fuel-rod assemblies. The coolants of different temperatures are mixed in the chimneys reducing the effect of stripping (hot-cold temperature fluctuations) on the remainder of the upper internals which are composed typically of AISI-304 or AISI-316 stainless steel.

  3. Reactor safety method

    DOEpatents

    Vachon, Lawrence J.

    1980-03-11

    This invention relates to safety means for preventing a gas cooled nuclear reactor from attaining criticality prior to start up in the event the reactor core is immersed in hydrogenous liquid. This is accomplished by coating the inside surface of the reactor coolant channels with a neutral absorbing material that will vaporize at the reactor's operating temperature.

  4. Further Development of Crack Growth Detection Techniques for US Test and Research Reactors

    SciTech Connect

    Kohse, Gordon; Carpenter, David M.; Ostrovsky, Yakov; Joseph Palmer, A.; Teysseyre, Sebastien P.; Davis, Kurt L.; Rempe, Joy L.

    2015-07-01

    One of the key issues facing Light Water Reactors (LWRs) in extending lifetimes beyond 60 years is characterizing the combined effect of irradiation and water chemistry on material degradation and failure. Irradiation Assisted Stress Corrosion Cracking (IASCC), in which a crack propagates in a susceptible material under stress in an aggressive environment, is a mechanism of particular concern. Full understanding of IASCC depends on real time crack growth data acquired under relevant irradiation conditions. Techniques to measure crack growth in actively loaded samples under irradiation have been developed outside the US - at the Halden Boiling Water Reactor, for example. Several types of IASCC tests have also been deployed at the MITR, including passively loaded crack growth measurements and actively loaded slow strain rate tests. However, there is not currently a facility available in the US to measure crack growth on actively loaded, pre-cracked specimens in LWR irradiation environments. A joint program between the Idaho National Laboratory (INL) and the Massachusetts Institute of Technology (MIT) Nuclear Reactor Laboratory (NRL) is currently underway to develop and demonstrate such a capability for US test and research reactors. Based on the Halden design, the samples will be loaded using miniature high pressure bellows and a compact loading mechanism, with crack length measured in real time using the switched Direct Current Potential Drop (DCPD) method. The basic design and initial mechanical testing of the load system and implementation of the DCPD method have been previously reported. This paper presents the results of initial autoclave testing at INL and the adaptation of the design for use in the high pressure, high temperature water loop at the MITR 6 MW research reactor, where an initial demonstration is planned in mid-2015. Materials considerations for the high pressure bellows are addressed. Design modifications to the loading mechanism required by the

  5. Nuclear reactor

    DOEpatents

    Thomson, Wallace B.

    2004-03-16

    A nuclear reactor comprising a cylindrical pressure vessel, an elongated annular core centrally disposed within and spaced from the pressure vessel, and a plurality of ducts disposed longitudinally of the pressure vessel about the periphery thereof, said core comprising an annular active portion, an annular reflector just inside the active portion, and an annular reflector just outside the active a portion, said annular active portion comprising rectangular slab, porous fuel elements radially disposed around the inner reflector and extending the length of the active portion, wedge-shaped, porous moderator elements disposed adjacent one face of each fuel element and extending the length of the fuel element, the fuel and moderator elements being oriented so that the fuel elements face each other and the moderator elements do likewise, adjacent moderator elements being spaced to provide air inlet channels, and adjacent fuel elements being spaced to provide air outlet channels which communicate with the interior of the peripheral ducts, and means for introducing air into the air inlet channels which passes through the porous moderator elements and porous fuel elements to the outlet channel.

  6. The Capabilities and Limitation of Remote Visual Methods to Detect Service-Induced Cracks in Reactor Components

    SciTech Connect

    Cumblidge, Stephen E.; Doctor, Steven R.; Anderson, Michael T.

    2006-11-01

    Since 1977, the U.S. Nuclear Regulatory Commission (NRC) Office of Nuclear Regulatory Research has funded a multiyear program at the Pacific Northwest National Laboratory (PNNL) to evaluate the reliability and accuracy of nondestructive evaluation (NDE) techniques employed for inservice inspection (ISI). Recently, the U.S. nuclear industry proposed replacing current volumetric and/or surface examinations of certain components in commercial nuclear power plants, as required by ASME Boiler and Pressure Vessel Code Section XI, with a simpler visual testing (VT) method. The advantages of VT are that these tests generally involve much less radiation exposure and examination times than do volumetric examinations such as ultrasonic testing (UT). However, for industry to justify supplamenting volumetric metods with VT, and analysis of pertinent issues is needed to support the reliability of VT in determining the structural intefrity of reactor components. As piping and pressure vessel compoents in a nuclear power station are generally underwater and in high radiation field, they need to be examined by VT from a distance with radiation-hardened video systems. Remote visual testing has been used by nuclear utilities to find cracks in pressure vessel cladding in pressurized water reactors, for shrouds in boiling water reactors, and to investigate leaks in piping and reactor components. These visual tests are performed using a wide variety of procedures and equipment. The techniques for remote visual testing use submersible closed-circuit video cameras to examine reactor components and welds. PNNL has conducted a parametric study that examines the important variables that affect the effectiveness of a remote visual test. Tested variables include lighting techniques, camera resolution, camera movement, and magnification. PNNL has also conductrd a laboratory test using a commercial visual testing camera system to experimentally determine the ability of the camera system to

  7. Etude aerodynamique d'un jet turbulent impactant une paroi concave

    NASA Astrophysics Data System (ADS)

    LeBlanc, Benoit

    Etant donne la demande croissante de temperatures elevees dans des chambres de combustion de systemes de propulsions en aerospatiale (turbomoteurs, moteur a reaction, etc.), l'interet dans le refroidissement par jets impactant s'est vu croitre. Le refroidissement des aubes de turbine permet une augmentation de temperature de combustion, ce qui se traduit en une augmentation de l'efficacite de combustion et donc une meilleure economie de carburant. Le transfert de chaleur dans les au bages est influence par les aspects aerodynamiques du refroidissement a jet, particulierement dans le cas d'ecoulements turbulents. Un manque de comprehension de l'aerodynamique a l'interieur de ces espaces confinees peut mener a des changements de transfert thermique qui sont inattendus, ce qui augmente le risque de fluage. Il est donc d'interet pour l'industrie aerospatiale et l'academie de poursuivre la recherche dans l'aerodynamique des jets turbulents impactant les parois courbes. Les jets impactant les surfaces courbes ont deja fait l'objet de nombreuses etudes. Par contre des conditions oscillatoires observees en laboratoire se sont averees difficiles a reproduire en numerique, puisque les structures d'ecoulements impactants des parois concaves sont fortement dependantes de la turbulence et des effets instationnaires. Une etude experimentale fut realisee a l'institut PPRIME a l'Universite de Poitiers afin d'observer le phenomene d'oscillation dans le jet. Une serie d'essais ont verifie les conditions d'ecoulement laminaires et turbulentes, toutefois le cout des essais experimentaux a seulement permis d'avoir un apercu du phenomene global. Une deuxieme serie d'essais fut realisee numeriquement a l'Universite de Moncton avec l'outil OpenFOAM pour des conditions d'ecoulement laminaire et bidimensionnel. Cette etude a donc comme but de poursuivre l'enquete de l'aerodynamique oscillatoire des jets impactant des parois courbes, mais pour un regime d'ecoulement transitoire, turbulent

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

    SciTech Connect

    Baratta, A.J.

    1997-07-01

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

  9. 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).

  10. Tokamak reactor studies

    SciTech Connect

    Baker, C.C.

    1981-01-01

    This paper presents an overview of tokamak reactor studies with particular attention to commercial reactor concepts developed within the last three years. Emphasis is placed on DT fueled reactors for electricity production. A brief history of tokamak reactor studies is presented. The STARFIRE, NUWMAK, and HFCTR studies are highlighted. Recent developments that have increased the commercial attractiveness of tokamak reactor designs are discussed. These developments include smaller plant sizes, higher first wall loadings, improved maintenance concepts, steady-state operation, non-divertor particle control, and improved reactor safety features.

  11. Validation of standardized computer analyses for licensing evaluation/TRITON two-dimensional and three-dimensional models for light water reactor fuel

    SciTech Connect

    Bowman, S. M.; Gill, D. F.

    2006-07-01

    The isotopic depletion capabilities of the new Standardized Computer Analyses for Licensing Evaluation control module TRITON, coupled with ORIGEN-S, were evaluated using spent fuel assays from several commercial light water reactors with both standard and mixed-oxide fuel assemblies. Calculations were performed using the functional modules NEWT and KENO-VI. NEWT is a two-dimensional, arbitrary-geometry, discrete-ordinates transport code, and KENO-VI is a three-dimensional Monte Carlo transport code capable of handling complex three-dimensional geometries. To validate the codes and data used in depletion calculations, numerical predictions were compared with experimental measurements for a total of 29 samples taken from the Calvert Cliffs, Obrigheim, and San Onofre pressurized water reactors and the Gundremmingen boiling water reactor. Similar comparisons have previously been performed at the Oak Ridge National Laboratory for the one-dimensional SAS2H control module. The SAS2H, TRITON/KENO-VI, and TRITON/NEWT results were compared for corresponding samples. All analyses showed that TRITON/KENO-VI and TRITON/NEWT produced typically similar or better results than SAS2H. The calculations performed in this validation study demonstrate that the depletion capabilities of TRITON accurately model spent fuel depletion and decay. (authors)

  12. Hybrid plasmachemical reactor

    SciTech Connect

    Lelevkin, V. M. Smirnova, Yu. G.; Tokarev, A. V.

    2015-04-15

    A hybrid plasmachemical reactor on the basis of a dielectric barrier discharge in a transformer is developed. The characteristics of the reactor as functions of the dielectric barrier discharge parameters are determined.

  13. Attrition reactor system

    SciTech Connect

    Scott, Charles D.; Davison, Brian H.

    1993-01-01

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

  14. Attrition reactor system

    SciTech Connect

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

    1993-09-28

    A reactor vessel for reacting a solid particulate with a liquid reactant has a centrifugal pump in circulatory flow communication with the reactor vessel for providing particulate attrition, resulting in additional fresh surface where the reaction can occur. 2 figures.

  15. Period meter for reactors

    DOEpatents

    Rusch, Gordon K.

    1976-01-06

    An improved log N amplifier type nuclear reactor period meter with reduced probability for noise-induced scrams is provided. With the reactor at low power levels a sampling circuit is provided to determine the reactor period by measuring the finite change in the amplitude of the log N amplifier output signal for a predetermined time period, while at high power levels, differentiation of the log N amplifier output signal provides an additional measure of the reactor period.

  16. NEUTRONIC REACTOR POWER PLANT

    DOEpatents

    Metcalf, H.E.

    1962-12-25

    This patent relates to a nuclear reactor power plant incorporating an air-cooled, beryllium oxide-moderated, pebble bed reactor. According to the invention means are provided for circulating a flow of air through tubes in the reactor to a turbine and for directing a sidestream of the circu1ating air through the pebble bed to remove fission products therefrom as well as assist in cooling the reactor. (AEC)

  17. NUCLEAR REACTOR CONTROL SYSTEM

    DOEpatents

    Epler, E.P.; Hanauer, S.H.; Oakes, L.C.

    1959-11-01

    A control system is described for a nuclear reactor using enriched uranium fuel of the type of the swimming pool and other heterogeneous nuclear reactors. Circuits are included for automatically removing and inserting the control rods during the course of normal operation. Appropriate safety circuits close down the nuclear reactor in the event of emergency.

  18. Improved vortex reactor system

    DOEpatents

    Diebold, James P.; Scahill, John W.

    1995-01-01

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

  19. Advanced Test Reactor Tour

    SciTech Connect

    Miley, Don

    2011-01-01

    The Advanced Test Reactor at Idaho National Laboratory is the foremost nuclear materials test reactor in the world. This virtual tour describes the reactor, how experiments are conducted, and how spent nuclear fuel is handled and stored. For more information about INL research, visit http://www.facebook.com/idahonationallaboratory.

  20. Advanced Test Reactor Tour

    ScienceCinema

    Miley, Don

    2016-07-12

    The Advanced Test Reactor at Idaho National Laboratory is the foremost nuclear materials test reactor in the world. This virtual tour describes the reactor, how experiments are conducted, and how spent nuclear fuel is handled and stored. For more information about INL research, visit http://www.facebook.com/idahonationallaboratory.

  1. NEUTRONIC REACTOR SHIELDING

    DOEpatents

    Borst, L.B.

    1961-07-11

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

  2. High solids fermentation reactor

    DOEpatents

    Wyman, Charles E.; Grohmann, Karel; Himmel, Michael E.; Richard, Christopher J.

    1993-01-01

    A fermentation reactor and method for fermentation of materials having greater than about 10% solids. The reactor includes a rotatable shaft along the central axis, the shaft including rods extending outwardly to mix the materials. The reactor and method are useful for anaerobic digestion of municipal solid wastes to produce methane, for production of commodity chemicals from organic materials, and for microbial fermentation processes.

  3. High solids fermentation reactor

    DOEpatents

    Wyman, Charles E.; Grohmann, Karel; Himmel, Michael E.; Richard, Christopher J.

    1993-03-02

    A fermentation reactor and method for fermentation of materials having greater than about 10% solids. The reactor includes a rotatable shaft along the central axis, the shaft including rods extending outwardly to mix the materials. The reactor and method are useful for anaerobic digestion of municipal solid wastes to produce methane, for production of commodity chemicals from organic materials, and for microbial fermentation processes.

  4. Efficient Silicon Reactor

    NASA Technical Reports Server (NTRS)

    Bates, H. E.; Hill, D. M.; Jewett, D. N.

    1983-01-01

    High-purity silicon efficiently produced and transferred by continuous two-cycle reactor. New reactor operates in relatively-narrow temperature rate and uses large surfaces area to minimize heat expenditure and processing time in producing silicon by hydrogen reduction of trichlorosilane. Two cycles of reactor consists of silicon production and removal.

  5. A MELCOR Application to Two Light Water Reactor Nuclear Power Plant Core Melt Scenarios with Assumed Cavity Flooding Action

    SciTech Connect

    Martin-Fuertes, Francisco; Martin-Valdepenas, Juan Manuel; Mira, Jose; Sanchez, Maria Jesus

    2003-10-15

    The MELCOR 1.8.4 code Bottom Head package has been applied to simulate two reactor cavity flooding scenarios for when the corium material relocates to the lower-plenum region in postulated severe accidents. The applications were preceded by a review of two main physical models, which highly impacted the results. A model comparison to available bibliography models was done, which allowed some code modifications on selected default assumptions to be undertaken. First, the corium convective heat transfer to the wall when it becomes liquid was modified, and second, the default nucleate boiling regime curve in a submerged hemisphere was replaced by a new curve (and, to a much lesser extent, the critical heat flux curve was slightly varied).The applications were devoted to two prototypical light water reactor nuclear power plants, a 2700-MW(thermal) pressurized water reactor (PWR) and a 1381-MW(thermal) boiling water reactor (BWR). The main conclusions of the cavity flooding simulations were that the PWR lower-head survivability is extended although it is clearly not guaranteed, while in the BWR sequence the corium seems to be successfully arrested in the lower plenum.Three applications of the CFX 4.4 computational fluid dynamics code were carried out in the context of the BWR scenario to support the first modification of the aforementioned two scenarios for MELCOR.Finally, in the same BWR context, a statistic predictor of selected output parameters as a function of input parameters is presented, which provides reasonable results when compared to MELCOR full calculations in much shorter CPU processing times.

  6. Etude sur les tendons en materiaux composites et leur application aux ancrages postcontraints

    NASA Astrophysics Data System (ADS)

    Chennouf, Adil

    L'objectif general de la presente these est d'evaluer le comportement a l'arrachement et au fluage d'ancrages injectes constitues de tendons en materiaux composites afin d'etablir des recommandations plus appropriees et realistes pour le dimensionnement et la conception. Quatre types de tendons en materiaux composites, deux a base de fibres d'aramide et deux a base de fibres de carbone, ont ete utilises dans l'etude. Les travaux de recherche de cette these ont porte notamment sur: (I) Une caracterisation physique et mecanique des tendons en materiaux composites utilises dans l'etude. (II) Une etude en laboratoire sur les coulis de scellement. La premiere etape de cette etude a concerne le developpement d'un coulis de scellement performant adapte aux tendons en materiaux composites et a differentes situations d'injection. La seconde etape a traite des essais de caracterisations physique et mecanique du coulis de scellement developpe comparativement a trois coulis de scellement usuels d'un meme rapport E/L de 0,4. (III) Une etude sur des modeles reduits d'ancrages injectes. (IV) Une etude sur des modeles d'ancrages a grande echelle. La synthese de ces etudes a permis d'enoncer les principales conclusions suivantes: (1) Les valeurs moyennes des charges de rupture des tendons en materiaux composites ont ete de 1% a 29% superieures a celles specifiees par les manufacturiers. (2) L'etude sur les coulis de scellement a permis le developpement de coulis de ciment repondant aux criteres fixes, soient une grande stabilite, une bonne fluidite, une legere expansion et de bonnes caracteristiques mecaniques. (3) Les tendons en materiaux composites ont montre des contraintes d'adherence maximum superieures a celles des tendons en acier. (4) Le type de fibre, la configuration et le fini de surface des tendons en materiaux composites gouvernent leur resistance a l'adherence. (5) L'introduction de sable et d'autres ajouts comme les fines de silice et la poudre d'aluminium au coulis

  7. Environmentally assisted cracking in light water reactors annual report January - December 2005.

    SciTech Connect

    Alexandreanu, B.; Chen, Y.; Chopra, O. K.; Chung, H. M.; Gruber, E. E.; Shack, W. J.; Soppet, W. K.

    2007-08-31

    This report summarizes work performed from January to December 2005 by Argonne National Laboratory on fatigue and environmentally assisted cracking in light water reactors (LWRs). Existing statistical models for estimating the fatigue life of carbon and low-alloy steels and austenitic stainless steels (SSs) as a function of material, loading, and environmental conditions were updated. Also, the ASME Code fatigue adjustment factors of 2 on stress and 20 on life were critically reviewed to assess the possible conservatism in the current choice of the margins. An approach, based on an environmental fatigue correction factor, for incorporating the effects of LWR environments into ASME Section III fatigue evaluations is discussed. The susceptibility of austenitic stainless steels and their welds to irradiation-assisted stress corrosion cracking (IASCC) is being evaluated as a function of the fluence level, water chemistry, material chemistry, and fabrication history. For this task, crack growth rate (CGR) tests and slow strain rate tensile (SSRT) tests are being conducted on various austenitic SSs irradiated in the Halden boiling water reactor. The SSRT tests are currently focused on investigating the effects of the grain boundary engineering process on the IASCC of the austenitic SSs. The CGR tests were conducted on Type 316 SSs irradiated to 0.45-3.0 dpa, and on sensitized Type 304 SS and SS weld heat-affected-zone material irradiated to 2.16 dpa. The CGR tests on materials irradiated to 2.16 dpa were followed by a fracture toughness test in a water environment. The effects of material composition, irradiation, and water chemistry on growth rates are discussed. The susceptibility of austenitic SS core internals to IASCC and void swelling is also being evaluated for pressurized water reactors. Both SSRT tests and microstructural examinations are being conducted on specimens irradiated in the BOR-60 reactor in Russia to doses up to 20 dpa. Crack growth rate data

  8. Reactor vessel support system

    DOEpatents

    Golden, Martin P.; Holley, John C.

    1982-01-01

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

  9. Nuclear reactor overflow line

    DOEpatents

    Severson, Wayne J.

    1976-01-01

    The overflow line for the reactor vessel of a liquid-metal-cooled nuclear reactor includes means for establishing and maintaining a continuous bleed flow of coolant amounting to 5 to 10% of the total coolant flow through the overflow line to prevent thermal shock to the overflow line when the reactor is restarted following a trip. Preferably a tube is disposed concentrically just inside the overflow line extending from a point just inside the reactor vessel to an overflow tank and a suction line is provided opening into the body of liquid metal in the reactor vessel and into the annulus between the overflow line and the inner tube.

  10. Thorium fueled reactor

    NASA Astrophysics Data System (ADS)

    Sipaun, S.

    2017-01-01

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

  11. EPA Statement on Flint Water Main Break, Boil Water Order

    EPA Pesticide Factsheets

    FLINT, MICH. -- The U.S. Environmental Protection Agency is working closely with the City of Flint and Michigan Department of Environmental Quality on the recent water main break and boil order. After the water transmission line broke on Feb. 9, EPA coordi

  12. Singing as a Therapeutic Agent, inThe Etude, 1891-1949.

    PubMed

    Hunter

    1999-01-01

    The Etude music magazine, founded by Theodore Presser, was one of a number of popular music magazines published in the years prior to the establishment of the music therapy profession in 1950. During its publication run from 1883 to 1957, over 100 music therapy related articles appeared, including 13 on the health benefits of singing published between 1891 and 1949. Written by authors with diverse backgrounds, such as the famous Battle Creek, Michigan physician John Harvey Kellogg and Boston music critic Louis C. Elson, the articles contained consistent and adamant support regarding the health benefits of singing. The advantages described were both physical and psychological, and were recommended prophylactically for well persons and therapeutically for ill persons. Although the articles varied in perspective, from philosophical to theoretical to pedagogical, there is a consistent holistic medicine theme that appeared almost ahead of its time and no doubt linked to the push for vocal music education in that era. The importance of The Etude in promulgating ideas that helped shape the early practice of music therapy should not be underestimated. For much of its publication run The Etude was the largest music periodical in print, reaching its peak circulation of 250,000 copies per month in 1924.

  13. 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.

  14. Spinning fluids reactor

    DOEpatents

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

    2012-11-20

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

  15. 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.

  16. Reactor water cleanup system

    DOEpatents

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

    1994-12-20

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

  17. Neutron fluxes in test reactors

    SciTech Connect

    Youinou, Gilles Jean-Michel

    2017-01-01

    Communicate the fact that high-power water-cooled test reactors such as the Advanced Test Reactor (ATR), the High Flux Isotope Reactor (HFIR) or the Jules Horowitz Reactor (JHR) cannot provide fast flux levels as high as sodium-cooled fast test reactors. The memo first presents some basics physics considerations about neutron fluxes in test reactors and then uses ATR, HFIR and JHR as an illustration of the performance of modern high-power water-cooled test reactors.

  18. Radioactive Fission Product Release from Defective Light Water Reactor Fuel Elements

    SciTech Connect

    Konyashov, Vadim V.; Krasnov, Alexander M.

    2002-04-15

    Results are provided of the experimental investigation of radioactive fission product (RFP) release, i.e., krypton, xenon, and iodine radionuclides from fuel elements with initial defects during long-term (3 to 5 yr) irradiation under low linear power (5 to 12 kW/m) and during special experiments in the VK-50 vessel-type boiling water reactor.The calculation model for the RFP release from the fuel-to-cladding gap of the defective fuel element into coolant was developed. It takes into account the convective transport in the fuel-to-cladding gap and RFP sorption on the internal cladding surface and is in good agreement with the available experimental data. An approximate analytical solution of the transport equation is given. The calculation dependencies of the RFP release coefficients on the main parameters such as defect size, fuel-to-cladding gap, temperature of the internal cladding surface, and radioactive decay constant were analyzed.It is shown that the change of the RFP release from the fuel elements with the initial defects during long-term irradiation is, mainly, caused by fuel swelling followed by reduction of the fuel-to-cladding gap and the fuel temperature. The calculation model for the RFP release from defective fuel elements applicable to light water reactors (LWRs) was developed. It takes into account the change of the defective fuel element parameters during long-term irradiation. The calculation error according to the program does not exceed 30% over all the linear power change range of the LWR fuel elements (from 5 to 26 kW/m)

  19. Environmental distribution and long-term dispersion of reactor /sup 14/CO/sub 2/ around two German nuclear power plants

    SciTech Connect

    Levin, I.; Kromer, B.; Barabas, M.; Muennich, K.O.

    1988-02-01

    Carbon-14 data on atmospheric CO/sub 2/ as well as on plant material (tree leaves and wheat) from the vicinity of two German boiling water reactors (Philippsburg and Isar/Ohu) are reported. Atmospheric CO/sub 2/ samples taken routinely with an integration time of one or two weeks 1.75 km downwind of the Philippsburg reactor (900 MW electrical power) show a maximum /sup 14/C excess concentration of delta /sup 14/C (excess) = 300 +/- 7%, corresponding to 12.7 mBq m-3 (STP air). The long-term average excess amounts to delta /sup 14/C (excess) = 47 +/- 3%, corresponding to 2.0 mBq m-3 (STP air). The concentrations observed with plant material at the same sampling site range between delta /sup 14/C (excess) = 0% and 120%, corresponding to 0 and 27 mBq (g carbon)-1. With the meteorological dispersion parameters actually measured at the nuclear power plants, the dispersion factors for the various sampling sites and for the individual periods of sampling were calculated on the basis of a one-dimensional Gaussian plume model. With the observed /sup 14/C excess concentrations and the dispersion factor, a theoretical (i.e. calculated) reactor /sup 14/C source strength is then determined. For the Philippsburg reactor, which is situated in the flat Rhine valley, the theoretical and the observed yearly mean /sup 14/C emissions compare rather well (within a factor of 2). A significant systematical deviation from the model was found in the concentration decrease with source distance: the decrease predicted between the 1.75-km and 3.25-km distances is steeper than actually observed. The /sup 14/C excess concentrations found in tree leaves around the Isar/Ohu reactor (907 MW electrical power) at 1-2 km distance fall into the same range as observed at Philippsburg.

  20. Etude theorique des fluctuations structurales dans les composes organiques a dimensionnalite reduite

    NASA Astrophysics Data System (ADS)

    Dumoulin, Benoit

    Les systemes a dimensionnalite reduite constituent maintenant une branche entiere de la physique de la matiere condensee. Cette derniere s'est developpee rapidement au cours des dernieres annees, avec la decouverte des materiaux organiques qui presentent, justement, des proprietes physiques fortement anisotropes. Cette these presente une etude en trois parties de plusieurs composes organiques qui, bien que tres differents du point de vue de leurs compositions chimiques et de leurs proprietes physiques a haute temperature, subissent tous une instabilite structurale a tres basse temperature. De plus, dans chacun des cas, l'instabilite structurale est precedee d'un important regime fluctuatif a partir duquel les proprietes physiques changent de maniere significative. Notre etude suit un ordre chronologique inverse puisque nous nous attardons en premier lieu au cas de composes recemment decouverts: les composes de la famille des (BCPTTF)2X (X = PF6 , AsF6). Ces derniers sont des isolants magnetiques a la temperature ambiante et subissent une instabilite structurale de type spin-Peierls a une temperature appelee TSP. En particulier, nous nous interessons a l'etude des proprietes physiques de ces systemes dans le regime fluctuatif, qui precede cette instabilite. Notre etude theorique nous permet de comprendre en detail comment ces systemes s'approchent de l'instabilite struturale. Dans la seconde partie de cette these, nous etudions le regime fluctuatif (pre-transitionnel) observe experimentalement dans le compose de (TMTTF)2PF6. Ce compose organique, dont la structure s'apparente aux sels de Bechgaard, subit une instabilite de type spin-Peierls a une temperature T SP = 19K. Bien que ce compose possede la particularite d'etre un bon conducteur a la temperature ambiante, il subit une transition de type Mott-Hubbard a une temperature Trho ≈ 220K et devient alors un isolant magnetique, analogue aux composes de la famille des (BCPTTF)2X. Le regime fluctuatif precedant l

  1. A STRONGLY COUPLED REACTOR CORE ISOLATION COOLING SYSTEM MODEL FOR EXTENDED STATION BLACK-OUT ANALYSES

    SciTech Connect

    Zhao, Haihua; Zhang, Hongbin; Zou, Ling; Martineau, Richard Charles

    2015-03-01

    The reactor core isolation cooling (RCIC) system in a boiling water reactor (BWR) provides makeup cooling water to the reactor pressure vessel (RPV) when the main steam lines are isolated and the normal supply of water to the reactor vessel is lost. The RCIC system operates independently of AC power, service air, or external cooling water systems. The only required external energy source is from the battery to maintain the logic circuits to control the opening and/or closure of valves in the RCIC systems in order to control the RPV water level by shutting down the RCIC pump to avoid overfilling the RPV and flooding the steam line to the RCIC turbine. It is generally considered in almost all the existing station black-out accidents (SBO) analyses that loss of the DC power would result in overfilling the steam line and allowing liquid water to flow into the RCIC turbine, where it is assumed that the turbine would then be disabled. This behavior, however, was not observed in the Fukushima Daiichi accidents, where the Unit 2 RCIC functioned without DC power for nearly three days. Therefore, more detailed mechanistic models for RCIC system components are needed to understand the extended SBO for BWRs. As part of the effort to develop the next generation reactor system safety analysis code RELAP-7, we have developed a strongly coupled RCIC system model, which consists of a turbine model, a pump model, a check valve model, a wet well model, and their coupling models. Unlike the traditional SBO simulations where mass flow rates are typically given in the input file through time dependent functions, the real mass flow rates through the turbine and the pump loops in our model are dynamically calculated according to conservation laws and turbine/pump operation curves. A simplified SBO demonstration RELAP-7 model with this RCIC model has been successfully developed. The demonstration model includes the major components for the primary system of a BWR, as well as the safety

  2. High energy reactor neutrinos

    NASA Astrophysics Data System (ADS)

    Raper, Neill

    We present the first measurement of a nonzero reactor neutrino flux with energies above 8 MeV. Measurements are taken with the Daya Bay Reactor Neutrino Experiments detectors, using the Guangdong Nuclear Power Station as a source. Disagreement between data and theory regarding rate and shape of reactor neutrino spectra have made the need for direct measurement clear. Data are especially useful at high energies, where far fewer isotopes contribute. Neutrino candidates are correlated to reactor power and reactor power is extrapolated to zero in order to separate neutrino events from background. We find evidence of reactor neutrinos up to ˜12.5 MeV at 1.92 sigma above 0 and include a survey of isotopes likely to be contributing neutrinos in this energy range.

  3. 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.

  4. Improved vortex reactor system

    DOEpatents

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

    1995-05-09

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

  5. The Integral Fast Reactor

    SciTech Connect

    Chang, Y.I.

    1988-01-01

    The Integral Fast Reactor (IFR) is an innovative liquid metal reactor concept being developed at Argonne National Laboratory. It seeks to specifically exploit the inherent properties of liquid metal cooling and metallic fuel in a way that leads to substantial improvements in the characteristics of the complete reactor system. This paper describes the key features and potential advantages of the IFR concept, with emphasis on its safety characteristics. 3 refs., 4 figs., 1 tab.

  6. NEUTRONIC REACTOR CONTROL

    DOEpatents

    Dreffin, R.S.

    1959-12-15

    A control means for a nuclear reactor is described. Particularly a device extending into the active portion of the reactor consisting of two hollow elements coaxially disposed and forming a channel therebetween, the cross sectional area of the channel increasing from each extremity of the device towards the center thereof. An element of neutron absorbing material is slidably positionable within the inner hollow element and a fluid reactor poison is introduced into the channel defined by the two hollow elements.

  7. NEUTRONIC REACTOR SYSTEM

    DOEpatents

    Goett, J.J.

    1961-01-24

    A system is described which includes a neutronic reactor containing a dispersion of fissionable material in a liquid moderator as fuel and a conveyor to which a portion of the dispersion may be passed and wherein the self heat of the slurry evaporates the moderator. Means are provided for condensing the liquid moderator and returning it to the reactor and for conveying the dried fissionable material away from the reactor.

  8. FLOW SYSTEM FOR REACTOR

    DOEpatents

    Zinn, W.H.

    1963-06-11

    A reactor is designed with means for terminating the reaction when returning coolant is below a predetermined temperature. Coolant flowing from the reactor passes through a heat exchanger to a lower reservoir, and then circulates between the lower reservoir and an upper reservoir before being returned to the reactor. Means responsive to the temperature of the coolant in the return conduit terminate the chain reaction when the temperature reaches a predetermined minimum value. (AEC)

  9. University Reactor Sharing Program

    SciTech Connect

    Dr. W.D. Reece

    1999-09-01

    The University Reactor Sharing Program provides funding for reactor experimentation to institutions that do not normally have access to a research reactor. Research projects supported by the program include items such as dating geological material to producing high current super conducting magnets. The funding also gives small colleges and universities the opportunity to use the facility for teaching courses in nuclear processes; specifically neutron activation analysis and gamma spectroscopy.

  10. Pressurized fluidized bed reactor

    DOEpatents

    Isaksson, Juhani

    1996-01-01

    A pressurized fluid bed reactor power plant includes a fluidized bed reactor contained within a pressure vessel with a pressurized gas volume between the reactor and the vessel. A first conduit supplies primary gas from the gas volume to the reactor, passing outside the pressure vessel and then returning through the pressure vessel to the reactor, and pressurized gas is supplied from a compressor through a second conduit to the gas volume. A third conduit, comprising a hot gas discharge, carries gases from the reactor, through a filter, and ultimately to a turbine. During normal operation of the plant, pressurized gas is withdrawn from the gas volume through the first conduit and introduced into the reactor at a substantially continuously controlled rate as the primary gas to the reactor. In response to an operational disturbance of the plant, the flow of gas in the first, second, and third conduits is terminated, and thereafter the pressure in the gas volume and in the reactor is substantially simultaneously reduced by opening pressure relief valves in the first and third conduits, and optionally by passing air directly from the second conduit to the turbine.

  11. Remote Reactor Monitoring

    SciTech Connect

    Bernstein, Adam; Dazeley, Steve; Dobie, Doug; Marleau, Peter; Brennan, Jim; Gerling, Mark; Sumner, Matthew; Sweany, Melinda

    2014-10-21

    The overall goal of the WATCHMAN project is to experimentally demonstrate the potential of water Cerenkov antineutrino detectors as a tool for remote monitoring of nuclear reactors. In particular, the project seeks to field a large prototype gadolinium-doped, water-based antineutrino detector to demonstrate sensitivity to a power reactor at ~10 kilometer standoff using a kiloton scale detector. The technology under development, when fully realized at large scale, could provide remote near-real-time information about reactor existence and operational status for small operating nuclear reactors out to distances of many hundreds of kilometers.

  12. Pressurized fluidized bed reactor

    DOEpatents

    Isaksson, J.

    1996-03-19

    A pressurized fluid bed reactor power plant includes a fluidized bed reactor contained within a pressure vessel with a pressurized gas volume between the reactor and the vessel. A first conduit supplies primary gas from the gas volume to the reactor, passing outside the pressure vessel and then returning through the pressure vessel to the reactor, and pressurized gas is supplied from a compressor through a second conduit to the gas volume. A third conduit, comprising a hot gas discharge, carries gases from the reactor, through a filter, and ultimately to a turbine. During normal operation of the plant, pressurized gas is withdrawn from the gas volume through the first conduit and introduced into the reactor at a substantially continuously controlled rate as the primary gas to the reactor. In response to an operational disturbance of the plant, the flow of gas in the first, second, and third conduits is terminated, and thereafter the pressure in the gas volume and in the reactor is substantially simultaneously reduced by opening pressure relief valves in the first and third conduits, and optionally by passing air directly from the second conduit to the turbine. 1 fig.

  13. HOMOGENEOUS NUCLEAR POWER REACTOR

    DOEpatents

    King, L.D.P.

    1959-09-01

    A homogeneous nuclear power reactor utilizing forced circulation of the liquid fuel is described. The reactor does not require fuel handling outside of the reactor vessel during any normal operation including complete shutdown to room temperature, the reactor being selfregulating under extreme operating conditions and controlled by the thermal expansion of the liquid fuel. The liquid fuel utilized is a uranium, phosphoric acid, and water solution which requires no gus exhaust system or independent gas recombining system, thereby eliminating the handling of radioiytic gas.

  14. Membrane reactors at Degussa.

    PubMed

    Wöltinger, Jens; Karau, Andreas; Leuchtenberger, Wolfgang; Drauz, Karlheinz

    2005-01-01

    The review covers the development of membrane reactor technologies at Degussa for the synthesis of fine chemicals. The operation of fed-batch or continuous biocatalytic processes in the enzyme membrane reactor (EMR) is well established at Degussa. Degussa has experience of running EMRs from laboratory gram scale up to a production scale of several hundreds of tons per year. The transfer of the enzyme membrane reactor from biocatalysis to chemical catalysis in the chemzyme membrane reactor (CMR) is discussed. Various homogeneous catalysts have been investigated in the CMR, and the scope and limitation of this new technique is discussed.

  15. Spent fuel assembly hardware: Characterization and 10 CFR 61 classification for waste disposal: Volume 2, Calculated activity profiles of spent nuclear fuel assembly hardware for pressurized water reactors

    SciTech Connect

    Short, S.M.; Luksic, A.T.; Lotz, T.L.; Schutz, M.E.

    1989-06-01

    Consolidation of spent fuel is under active consideration as the US Department of Energy plans to dispose of spent fuel as required by the Nuclear Waste Policy Act of 1982. During consolidation, the fuel pins are removed from an intact fuel assembly and repackaged into a more compact configuration. After repackaging, approximately 30 kg of residual spent fuel assembly hardware per assembly remains that is also radioactive and requires disposal. Understanding the nature of this secondary waste stream is critical to designing a system that will properly handle, package, store, and dispose of the waste. This report present a methodology for estimating the radionuclide inventory in irradiated spent fuel hardware. Ratios are developed that allow the use of ORIGEN2 computer code calculations to be applied to regions that are outside the fueled region. The ratios are based on the analysis of samples of irradiated hardware from spent fuel assemblies. The results of this research are presented in three volumes. In Volume 1, the development of scaling factors that can be used with ORIGEN2 calculations to estimate activation of spent fuel assembly hardware is documented. The results from Laboratory analysis of irradiated spent-fuel hardware samples are also presented in Volume 1. In Volumes 2 and 3, the calculated flux profiles of spent nuclear fuel assemblies are presented for pressurized water reactors and boiling water reactors, respectively. The results presented in Volumes 2 and 3 were used to develop the scaling factors documented in Volume 1.

  16. 10 CFR Appendix J to Part 50 - Primary Reactor Containment Leakage Testing for Water-Cooled Power Reactors

    Code of Federal Regulations, 2012 CFR

    2012-01-01

    ... and feedwater piping and other systems which penetrate containment of direct-cycle boiling water power..., pressurized with air, nitrogen, or pneumatic fluid specified in the technical specifications or associated... pressure loss of the test chamber of the containment penetration pressurized with air, nitrogen,...

  17. REFLECTOR FOR NEUTRONIC REACTORS

    DOEpatents

    Fraas, A.P.

    1963-08-01

    A reflector for nuclear reactors that comprises an assembly of closely packed graphite rods disposed with their major axes substantially perpendicular to the interface between the reactor core and the reflector is described. Each graphite rod is round in transverse cross section at (at least) its interface end and is provided, at that end, with a coaxial, inwardly tapering hole. (AEC)

  18. NEUTRONIC REACTOR BURIAL ASSEMBLY

    DOEpatents

    Treshow, M.

    1961-05-01

    A burial assembly is shown whereby an entire reactor core may be encased with lead shielding, withdrawn from the reactor site and buried. This is made possible by a five-piece interlocking arrangement that may be easily put together by remote control with no aligning of bolt holes or other such close adjustments being necessary.

  19. N Reactor hydrogen control

    SciTech Connect

    Shuford, D.H.; Kripps, L.J.

    1988-08-01

    Following the accident at the Chernobyl nuclear power reactor in the Soviet Union, a number of reviews were conducted of the N Reactor. Hydrogen generation during postulates severe accidents and the possibility of resulting hydrogen deflagrations/detonations that could affect confinement integrity were issues raised in several reviews, along with recommendations for adding hydrogen mitigation features. To respond to these reviews, an N Reactor Safety Enhancement Program and a subsequent Accelerated Safety Enhancement Program were initiated to address all post-Chernobyl N Reactor review findings. The Safety Enhancement Program and Accelerated Safety Enhancement Program efforts involving hydrogen control included the following: Calculate the potential hydrogen source for a range of severe accidents at the N Reactor to establish an acceptable design basis for the hydrogen mitigation system; Analyze the N Reactor confinement hydrogen mixing capability to identify areas of concern and to the verify effectiveness of the hydrogen mitigation system; Select, design, and construct a hydrogen mitigation system to enhance the N Reactor capability to accommodate possible hydrogen generation from postulated severe accidents; Provide post-accident hydrogen monitoring as an operator aid in assessing confinement conditions. In additions, it was necessary to verify that incorporation of the hydrogen mitigation system had no adverse impact N Reactor safety (e.g., radiological consequence analyses). 77 refs., 25 figs., 10 tabs.

  20. Polymerization Reactor Engineering.

    ERIC Educational Resources Information Center

    Skaates, J. Michael

    1987-01-01

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

  1. The Integral Fast Reactor

    SciTech Connect

    Till, C.E.; Chang, Y.I. ); Lineberry, M.J. )

    1990-01-01

    Argonne National Laboratory, since 1984, has been developing the Integral Fast Reactor (IFR). This paper will describe the way in which this new reactor concept came about; the technical, public acceptance, and environmental issues that are addressed by the IFR; the technical progress that has been made; and our expectations for this program in the near term. 5 refs., 3 figs.

  2. Polymerization Reactor Engineering.

    ERIC Educational Resources Information Center

    Skaates, J. Michael

    1987-01-01

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

  3. 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.

  4. Status of French reactors

    SciTech Connect

    Ballagny, A.

    1997-08-01

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

  5. REACTOR FUEL SCAVENGING MEANS

    DOEpatents

    Coffinberry, A.S.

    1962-04-10

    A process for removing fission products from reactor liquid fuel without interfering with the reactor's normal operation or causing a significant change in its fuel composition is described. The process consists of mixing a liquid scavenger alloy composed of about 44 at.% plutoniunm, 33 at.% lanthanum, and 23 at.% nickel or cobalt with a plutonium alloy reactor fuel containing about 3 at.% lanthanum; removing a portion of the fuel and scavenger alloy from the reactor core and replacing it with an equal amount of the fresh scavenger alloy; transferring the portion to a quiescent zone where the scavenger and the plutonium fuel form two distinct liquid layers with the fission products being dissolved in the lanthanum-rich scavenger layer; and the clean plutonium-rich fuel layer being returned to the reactor core. (AEC)

  6. Reactor Operations Management Plan

    SciTech Connect

    Rice, P.D.

    1991-12-05

    The K-Reactor last operated in April 1988. At that time, K-Reactor was one of three operating reactors at the Savannah River Site (SRS). Following an incident in P-Reactor in August 1988, it was decided to discontinue SRS reactor operation and conduct an extensive program to upgrade operating practices and plant hardware prior to restart of any of the reactors. The K-reactor was the first of three reactors scheduled to resume production. At the present time, it is the only reactor with planned restart. WSRC assumed management of SRS on April 1, 1989. WSRC established the Safety Basis for Restart and a listing of the actions planned to satisfy the Safety Basis. In consultation with DOE, it was determined that proper management of the restart activities would require a single plan that integrated the numerous activities. The plan was entitled the Reactor Operations Management Plan and is referred to simply as the ROMP. The initial version of ROMP was produced in July of 1989. Subsequent modifications led to Revision 3 which was approved by DOE in May, 1990. Other changes were made in a formal change process, resulting in the latest version, Revision 5, being issued in October, 1990. The ROMP contains three key parts: first, the Restart Safety Basis; second, a description of the process for addressing new technical issues and a listing of the established workscope and the associated acceptance criteria; and three, a schedule for executing the work. I will discuss the first two areas, along with the closure process used to assure the intent of ROMP was met. The ROMP activities are all complete and I will not discuss schedule further.

  7. Evaluation of the tritium content in light water reactor control and absorber rods to obtain data for the fuel cycle backend

    SciTech Connect

    Bleier, A.; Neeb, K.H.; Gelfort, E.; Mischke, J.

    1986-08-01

    Tritium inventories and tritium distribution have been determined in boron glass absorber rods discharged from a pressurized water reactor first-cycle core and in spent boron carbide (B/sub 4/C) control rods from a boiling water reactor. The total tritium inventory in the boron glass absorber rods from the Stade nuclear reactor amounts to approx. =8.0 x 10/sup 10/ Bq (2.2 Ci) per rod. Of this, 99.6% was fixed in the boron glass itself and 0.4% in the Al/sub 2/O/sub 3/ pellets. The 4 x 10/sup -3/% fractions in the tube cladding and support pipe and the 1 x 10/sup -2/% fraction in the fill gas accounted for an insignificant part of the total tritium inventory of the rod. This experimentally determined tritium inventory was a factor of 5 larger than that suggested by the calculated estimate. The discrepancy between analyzed and calculated values can be explained by tritium formation from lithium impurities in the boron glass, where a 30-ppm lithium content would be adequate for this tritium inventory to be generated by the reaction /sup 6/Li(n,..cap alpha..)/sup 3/H. Evaluation of the B/sub 4/C control rods from the Lingen nuclear reactor after 3 yr of operation gave a 3.2 x 10/sup 10/Bq(0.85-Ci)tritium inventory per B/sub 4/C rod, while the total tritium inventory for a control rod assembly containing 60 B/sub 4/C rods was approx. =1.9 x 10/sup 12/ Bq (50 Ci). The tritium generated was essentially bound 100% in the B/sub 4/C, since the hulls contained only 6 x 10/sup -3/% and the fill gas only 2 x 10/sup -4/%.

  8. Contributions a L'etude de Dispositifs D'optique Integree

    NASA Astrophysics Data System (ADS)

    Touam, Tahar

    Cette these contient des contributions a l'etude de deux champs du vaste domaine de l'optique integree. A cet effet, nous avons divise notre travail en deux grandes parties:. Dans une premiere partie, nous traitons le probleme de la realisation d'une nouvelle classe de guides d'onde planaires utilisables dans le domaine de longueur d'onde de l'infrarouge moyen (infrarouge thermique), domaine ou l'apparition anticipee de fibres optiques a pertes extremement faibles rendraient fort interessante l'existence de tels guides d'onde planaires. Dans un premier temps, nous presentons une etude analytique originale d'une structure planaire a profil d'indice gradue, suivie d'une analyse d'un guide canal base sur cette structure. Dans un deuxieme temps, nous decrivons le procede de fabrication par pulverisation atomique d'un guide planaire forme d'arseniure de gallium (AsGa) sur du dioxyde de silicium (SiO_2 ), combinaison de materiau compatible avec l'infrarouge moyen. Finalement, nous presentons une etude de conception d'un reseau de surface destine a coupler la lumiere dans un tel guide, les autres methodes traditionnelles de couplage semblant peu appropriees aux environs de lambda = 10 mum. Dans une deuxieme partie, nous traitons le probleme de la jonction Y en optique integree, jonction qui soufre de pertes tres importantes des que l'angle d'ouverture devient interessant pour le concepteur de circuits integres optiques. L'analyse est basee sur la methode numerique dite BPM (Beam Propagation Method; methode de propagation du faisceau) qui fait l'objet d'un bref rappel. Nous poursuivons avec l'etude et l'optimisation d'une nouvelle jonction Y dont l'essence est l'utilisation du phenomene de diffraction a travers trois fentes de phase. Nous obtenons ainsi une tres bonne jonction, separant proprement le faisceau, a une ouverture de 10 degres. Finalement, nous faisons un rappel d'un profil d'indice dit "ideal" pour guides courbes et nous proposons l'utilisation de tels guides

  9. REACTOR BASE, SOUTHEAST CORNER. INTERIOR WILL CONTAIN REACTOR TANK, COOLING ...

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

    REACTOR BASE, SOUTHEAST CORNER. INTERIOR WILL CONTAIN REACTOR TANK, COOLING WATER PIPES, COOLING AIR DUCTS, AND SHIELDING. INL NEGATIVE NO. 776. Unknown Photographer, 10/1950 - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

  10. Nuclear reactor control column

    DOEpatents

    Bachovchin, Dennis M.

    1982-01-01

    The nuclear reactor control column comprises a column disposed within the nuclear reactor core having a variable cross-section hollow channel and containing balls whose vertical location is determined by the flow of the reactor coolant through the column. The control column is divided into three basic sections wherein each of the sections has a different cross-sectional area. The uppermost section of the control column has the greatest cross-sectional area, the intermediate section of the control column has the smallest cross-sectional area, and the lowermost section of the control column has the intermediate cross-sectional area. In this manner, the area of the uppermost section can be established such that when the reactor coolant is flowing under normal conditions therethrough, the absorber balls will be lifted and suspended in a fluidized bed manner in the upper section. However, when the reactor coolant flow falls below a predetermined value, the absorber balls will fall through the intermediate section and into the lowermost section, thereby reducing the reactivity of the reactor core and shutting down the reactor.

  11. Reactor Safety Research Programs

    SciTech Connect

    Edler, S. K.

    1981-07-01

    This document summarizes the work performed by Pacific Northwest Laboratory (PNL) from January 1 through March 31, 1981, for the Division of Reactor Safety Research within the U.S. Nuclear Regulatory Commission (NRC). Evaluations of nondestructive examination (NDE) techniques and instrumentation are reported; areas of investigation include demonstrating the feasibility of determining the strength of structural graphite, evaluating the feasibility of detecting and analyzing flaw growth in reactor pressure boundary systems, examining NDE reliability and probabilistic fracture mechanics, and assessing the integrity of pressurized water reactor (PWR) steam generator tubes where service-induced degradation has been indicated. Experimental data and analytical models are being provided to aid in decision-making regarding pipeto- pipe impacts following postulated breaks in high-energy fluid system piping. Core thermal models are being developed to provide better digital codes to compute the behavior of full-scale reactor systems under postulated accident conditions. Fuel assemblies and analytical support are being provided for experimental programs at other facilities. These programs include loss-ofcoolant accident (LOCA) simulation tests at the NRU reactor, Chalk River, Canada; fuel rod deformation, severe fuel damage, and postaccident coolability tests for the ESSOR reactor Super Sara Test Program, Ispra, Italy; the instrumented fuel assembly irradiation program at Halden, Norway; and experimental programs at the Power Burst Facility, Idaho National Engineering Laboratory (INEL). These programs will provide data for computer modeling of reactor system and fuel performance during various abnormal operating conditions.

  12. Slurry reactor design studies

    SciTech Connect

    Fox, J.M.; Degen, B.D.; Cady, G.; Deslate, F.D.; Summers, R.L. ); Akgerman, A. ); Smith, J.M. )

    1990-06-01

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

  13. Scaling and design of PUMA and SBWR

    SciTech Connect

    Ishii, M.; Revankar, S.T.; Dowlati, R.; Bertodano, M.L.; Babelli, I.; Wang, W.; Pokharna, H.; Ransom, V.H.; Viskanta, R.

    1994-10-01

    The General Electric Nuclear Energy (GE) has developed a new boiling water reactor called the Simplified Boiling Water Reactor (SBWR). Major differences between the current Boiling Water Reactors (BWRs) and the SBWR are in the simplification of the coolant circulation system and the implementation of passive emergency cooling systems. There are no recirculation pumps to drive the coolant in the vessel of the SBWR. The emergency core cooling and containment cooling systems do not have active pump-injected flows.

  14. NUCLEAR REACTOR FUEL SYSTEMS

    DOEpatents

    Thamer, B.J.; Bidwell, R.M.; Hammond, R.P.

    1959-09-15

    Homogeneous reactor fuel solutions are reported which provide automatic recombination of radiolytic gases and exhibit large thermal expansion characteristics, thereby providing stability at high temperatures and enabling reactor operation without the necessity of apparatus to recombine gases formed by the radiolytic dissociation of water in the fuel and without the necessity of liquid fuel handling outside the reactor vessel except for recovery processes. The fuels consist of phosphoric acid and water solutions of enriched uranium, wherein the uranium is in either the hexavalent or tetravalent state.

  15. NEUTRONIC REACTOR CONTROL

    DOEpatents

    Metcalf, H.E.

    1958-10-14

    Methods of controlling reactors are presented. Specifically, a plurality of neutron absorber members are adjustably disposed in the reactor core at different distances from the center thereof. The absorber members extend into the core from opposite faces thereof and are operated by motive means coupled in a manner to simultaneously withdraw at least one of the absorber members while inserting one of the other absorber members. This feature effects fine control of the neutron reproduction ratio by varying the total volume of the reactor effective in developing the neutronic reaction.

  16. Microfluidic electrochemical reactors

    SciTech Connect

    Nuzzo, Ralph G; Mitrovski, Svetlana M

    2011-03-22

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

  17. Nuclear reactor reflector

    DOEpatents

    Hopkins, R.J.; Land, J.T.; Misvel, M.C.

    1994-06-07

    A nuclear reactor reflector is disclosed that comprises a stack of reflector blocks with vertical water flow passages to cool the reflector. The interface between blocks is opposite support points for reactor fuel rods. Water flows between the reflector and the reactor barrel from passages in a bottom block. The top block contains a flange to limit this flow and the flange has a slot to receive an alignment pin that is welded to the barrel. The pin is held in the slot by two removable shims. Alignment bars extend the length of the stack in slots machined in each block when the stack is assembled. 12 figs.

  18. Nuclear reactor reflector

    DOEpatents

    Hopkins, Ronald J.; Land, John T.; Misvel, Michael C.

    1994-01-01

    A nuclear reactor reflector is disclosed that comprises a stack of reflector blocks with vertical water flow passages to cool the reflector. The interface between blocks is opposite support points for reactor fuel rods. Water flows between the reflector and the reactor barrel from passages in a bottom block. The top block contains a flange to limit this flow and the flange has a slot to receive an alignment pin that is welded to the barrel. The pin is held in the slot by two removable shims. Alignment bars extend the length of the stack in slots machined in each block when the stack is assembled.

  19. COOLED NEUTRONIC REACTOR

    DOEpatents

    Binner, C.R.; Wilkie, C.B.

    1958-03-18

    This patent relates to a design for a reactor of the type in which a fluid coolant is flowed through the active portion of the reactor. This design provides for the cooling of the shielding material as well as the reactor core by the same fluid coolant. The core structure is a solid moderator having coolant channels in which are disposed the fuel elements in rod or slug form. The coolant fluid enters the chamber in the shield, in which the core is located, passes over the inner surface of said chamber, enters the core structure at the center, passes through the coolant channels over the fuel elements and out through exhaust ducts.

  20. REACTOR CONTROL SYSTEM

    DOEpatents

    MacNeill, J.H.; Estabrook, J.Y.

    1960-05-10

    A reactor control system including a continuous tape passing through a first coolant passageway, over idler rollers, back through another parallel passageway, and over motor-driven rollers is described. Discrete portions of fuel or poison are carried on two opposed active sections of the tape. Driving the tape in forward or reverse directions causes both active sections to be simultaneously inserted or withdrawn uniformly, tending to maintain a more uniform flux within the reactor. The system is particularly useful in mobile reactors, where reduced inertial resistance to control rod movement is important.