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Sample records for tmi-2 reactor building

  1. TMI-2 and reactor safety

    International Nuclear Information System (INIS)

    Kato, W.Y.

    1982-01-01

    The accident on March 28, 1979 at the Three Mile Island Unit 2 (TMI-2) Nuclear Power Station located near Harrisburg, Pennsylvania has had a major impact upon the design, construction, operation, regulation and safety research of nuclear power plants. It also has had an impact on utility and regulatory organizations. This paper presents a review of the TMI-2 accident scenario, its consequences and the conclusions of various investigating groups. An overview of the nuclear licensing process, computer codes used in safety analysis, safety research, and current issues is provided, using as a basis the TMI-2 accident. Some of the current safety issues which are included in the proposed NRC rulemaking hearings on such subjects as safety goals, siting criteria, emergency plans, and degraded or damaged core analysis are discussed. (author)

  2. Psychological stress for alternatives of decontamination of TMI-2 reactor building atmosphere. Technical report

    International Nuclear Information System (INIS)

    Baum, A.; Gatchel, R.; Streufert, S.; Baum, C.S.; Fleming, R.

    1980-08-01

    The purpose of the report is to consider the nature and level of psychological stress that may be associated with each of several alternatives for decontamination. The report briefly reviews some of the literature on stress, response to major disaster or life stressors, provides opinion on each decontamination alternative, and considers possible mitigative actions to reduce psychological stress. The report concludes that any procedure that is adapted for the decontamination of the reactor building atmosphere will result in some psychological stress. The stress, however, should abate as contamination is reduced and uncertainty is diminished. The advantages of the purge alternative are the rapid completion of the decontamination and the consequent elimination of future uncontrolled release. Severe stress effects are less likely if the duration of stressor exposure is reduced, if the feeling of public control is increased and if the degree of perceived safety is increased. The long delays, continued uncertainty, and possibility of uncontrolled release that characterize the other alternatives may offset the perception that they are safer. In addition, chronic stress could be a consequence of long delays and continued uncertainty

  3. TMI-2 reactor vessel head removal

    International Nuclear Information System (INIS)

    Bengel, P.R.; Smith, M.D.; Estabrook, G.A.

    1985-09-01

    This report describes the safe removal and storage of the Three Mile Island Unit 2 (TMI-2) reactor vessel head. The head was removed in July 1984 to permit the removal of the plenum and the reactor core, which were damaged during the 1979 accident. From July 1982, plans and preparations were made using a standard head removal procedure modified by the necessary precautions and changes to account for conditions caused by the accident. After data acquisition, equipment and structure modifications, and training, the head was safely removed and stored; and the internals indexing fixture and a work platform were installed on top of the vessel. Dose rates during and after the operation were lower than expected; lessons were learned from the operation which will be applied to the continuing fuel removal operations activities

  4. TMI-2 reactor vessel head removal

    Energy Technology Data Exchange (ETDEWEB)

    Bengel, P.R.; Smith, M.D.; Estabrook, G.A.

    1985-09-01

    This report describes the safe removal and storage of the Three Mile Island Unit 2 (TMI-2) reactor vessel head. The head was removed in July 1984 to permit the removal of the plenum and the reactor core, which were damaged during the 1979 accident. From July 1982, plans and preparations were made using a standard head removal procedure modified by the necessary precautions and changes to account for conditions caused by the accident. After data acquisition, equipment and structure modifications, and training, the head was safely removed and stored; and the internals indexing fixture and a work platform were installed on top of the vessel. Dose rates during and after the operation were lower than expected; lessons were learned from the operation which will be applied to the continuing fuel removal operations activities.

  5. Methods for eluting radiocesium from zeolite ion exchange material in a column in the TMI-2 reactor containment building

    International Nuclear Information System (INIS)

    Knauer, J.B.; Campbell, D.O.; Collins, E.D.; King, L.J.

    1982-07-01

    Two alternative procedures were evaluated at Oak Ridge National Laboratory for potential use in eluting the radiocesium from Linde Ionsiv IE-95 zeolite in the pushcart ion exchange column in the TMI-2 containment building. The elution mechanism was iosotopic exchange of the radiocesium with stable cesium. Small zeolite ion exchange columns that had been loaded during ORNL tests of the Submerged Demineralizer System (SDS) flowsheet were eluted during these tests. One column was eluted using 0.25 M CsNO 3 , and a second column was eluted using 0.25 M CsH 2 BO 3 . Both eluent solutions were effective for removing the cesium. The 0.25 CsNO 3 eluent removed approx. 91% of the 137 Cs in 20 bed volumes and approx. 92% in 37.5 bed volumes. The 0.25 M CsH 2 BO 3 eluent removed approx. 82% of the 137 Cs in 20 bed volumes and approx. 85% in 40 bed volumes. In both cases, the radiation levels on the columns were reduced by a factor of approx. 30

  6. TMI-2 reactor vessel head removal

    International Nuclear Information System (INIS)

    Bengel, P.R.; Smith, M.D.; Estabrook, G.A.

    1984-12-01

    This report describes the safe removal and storage of the Three Mile Island Unit 2 reactor vessel head. The head was removed in July 1984 to permit the removal of the plenum and the reactor core, which were damaged during the 1979 accident. From July 1982, plans and preparations were made using a standard head removal procedure modified by the necessary precautions and changes to account for conditions caused by the accident. After data acquisition, equipment and structure modifications, and training the head was safely removed and stored and the internals indexing fixture and a work platform were installed on top of the vessel. Dose rates during and after the operation were lower than expected; lessons were learned from the operation which will be applied to the continuing fuel removal operations activities

  7. TMI-2 reactor vessel head removal

    Energy Technology Data Exchange (ETDEWEB)

    Bengel, P.R.; Smith, M.D.; Estabrook, G.A.

    1984-12-01

    This report describes the safe removal and storage of the Three Mile Island Unit 2 reactor vessel head. The head was removed in July 1984 to permit the removal of the plenum and the reactor core, which were damaged during the 1979 accident. From July 1982, plans and preparations were made using a standard head removal procedure modified by the necessary precautions and changes to account for conditions caused by the accident. After data acquisition, equipment and structure modifications, and training the head was safely removed and stored and the internals indexing fixture and a work platform were installed on top of the vessel. Dose rates during and after the operation were lower than expected; lessons were learned from the operation which will be applied to the continuing fuel removal operations activities.

  8. What actually happened at TMI-2

    International Nuclear Information System (INIS)

    Duco, J.

    1989-08-01

    The 1979 Three Mile Island Unit 2 (TMI-2; USA) accident is evaluated. The reactor core is damaged and the released fission products contaminated the reactor building and the secondary buildings. The radiation dose on the TMI-2 personnel is about 4694 man-rem, from 1979 to 1987. The main steps of the accident and the TMI-2 decontamination actions are described. The accident showed the importance of the reactor vessel's water re-filling as soon as possible during accident [fr

  9. TMI-2 reactor vessel and balance of plant status

    International Nuclear Information System (INIS)

    Kuehn, G.A.

    1990-01-01

    In the fall of 1988 a corporate decision was made which concentrated effort on support of reactor vessel defueling and minimized activity in balance-of-plant areas. The auxiliary and fuel handling building are in a safe/stable state but final preparations for monitored storage won't be pursued until defueling and fuel shipping are complete. In addition to dispositioning fuel, the project is actively preparing for disposal of the Accident Generated Water (2.3 million gallons) by evaporation

  10. TMI-2 [Three Mile Island Unit 2] reactor vessel head and plenum removal

    International Nuclear Information System (INIS)

    Hultman, C.W.; Jackson, R.W.

    1988-01-01

    As a result of the damage to the Three Mile Island Unit 2 (TMI-2) core during the accident in March 1979, removal of the reactor vessel head and plenum in the normal manner was judged to be impossible. In June 1982, a task force was formed to develop a plan for removing the TMI-2 reactor vessel head. The plan proposed removing the head using the standard head removal procedure in conjunction with the necessary precautions, changes, and preparations required for potential problems. This included the potential for both higher than normal radiation levels and airborne radioactive contamination. In addition, the plan specified that the plant be left either in a condition to proceed with plenum and fuel removal immediately after head removal or in a safe long-term lay-up condition. The basic plan consisting of the described major steps is discussed

  11. TMI-2 pyrophoricity studies

    International Nuclear Information System (INIS)

    Baston, V.F.; Austin, W.E.; Hoffstetter, K.J.; Owen, D.E.

    1984-11-01

    This report summarizes an extensive literature review and experimental evaluation of the pyrophoric potential of debris and material from within the TMI-2 reactor system. The literature on pyrophoricity recommends educated caution during handling of the TMI-2 core debris because of the presence of zirconium, zirconium alloys, and the potential for some oxidizable zirconium compounds. Laboratory testing of zirconium-bearing TMI-2 core debris specimens obtained from locations within the reactor vessel show no pyrophoric potential. These data support TMI-2 accident analyses that have concluded that most of the zirconium particulates in the core were oxidized by the high temperature steam environment during the accident. In addition, the presence of other constituents in the core debris effectively dilutes any oxidizable zirconium, thereby rendering sustained propagation of any pyrophoric events unlikely. Consequently, a reactor defueling scheme has been developed for TMI-2 which acknowledges that, while complete understanding of the zirconium oxidation state is impossible, the safe handling techniques to be used during defueling minimize the concern for any pyrophoric event

  12. TMI-2 core examination

    International Nuclear Information System (INIS)

    Hobbins, R.R.; MacDonald, P.E.; Owen, D.E.

    1983-01-01

    The examination of the damaged core at the Three Mile Island Unit 2 (TMI-2) reactor is structured to address the following safety issues: fission product release, transport, and deposition; core coolability; containment integrity; and recriticality during severe accidents; as well as zircaloy cladding ballooning and oxidation during so-called design basis accidents. The numbers of TMI-2 components or samples to be examined, the priority of each examination, the safety issue addressed by each examination, the principal examination techniques to be employed, and the data to be obtained and the principal uses of the data are discussed in this paper

  13. TMI-2 analysis exercise

    International Nuclear Information System (INIS)

    1990-01-01

    On March 28, 1979 an accident at the Three Mile Island Unit 2 (TMI-2) nuclear power plant led to melting of approximately 50% of the reactor's core. Since the TMI-2 accident has not been duplicated in an experimental facility, TMI-2 provides a unique opportunity to assess the capability of our severe accident analysis methods to simulate an accident in a full scale nuclear power plant. In this regard the OECD Nuclear Energy Agency (NEA) in collaboration with the US Department of Energy (DOE), has established a Joint Task Group to analyze various periods of the accident in an effort to benchmark severe accident computer codes. Participants in the analysis exercise and contributors to this report are provided and the results and conclusions that may be drawn from these results of the benchmark analyses are documented in this report. 12 refs., 10 figs., 5 tabs

  14. Chemical and X-ray diffraction analysis on selected samples from the TMI-2 reactor core

    International Nuclear Information System (INIS)

    Kleykamp, H.; Pejsa, R.

    1991-05-01

    Selected samples from different positions of the damaged TMI-2 reactor core were investigated by X-ray microanalysis and X-ray diffraction. The measurements yield the following resolidified phases after cooling: Cd and In depleted Ag absorber material, intermetallic Zr-steel compounds, fully oxidized Zircaloy, UO 2 -ZrO 2 solid solutions and their decomposed phases, and Fe-Al-Cr-Zr spinels. The composition of the phases and their lattice parameters as well as the eutectic and monotectic character can serve as indicators of local temperatures of the core. The reaction sequences are estimated from the heterogeneous equilibria of these phases. The main conclusions are: (1) Liquefaction onset is locally possible by Inconel-Zircaloy and steel-Zircaloy reactions of spacers and absorber guide tubes at 930deg C. However, increased rates of dissolution occur above 1200deg C. (2) UO 2 dissolution in the Inconel-steel-Zircaloy melt starts at 1300deg C with increased rates above 1900deg C. (3) Fuel temperatures in the core centre are increased above 2550deg C, liquid (U,Zr)O 2 is generated. (4) Square UO 2 particles are reprecipitated from the Incoloy-steel-Zircaloy-UO 2 melt during cooling, the remaining metallic melt is oxygen poor; two types of intermetallic phases are formed. (5) Oxidized Fe and Zr and Al 2 O 3 from burnable absorber react to spinels which form a low melting eutectic with the fuel at 1500deg C. The spinel acts as lubricant for fuel transport to the lower reactor plenum above 1500deg C. (6) Ruthenium (Ru-106) is dissolved in the steel phase, antimony (Sb-125) in the α-Ag absorber during liquefaction. (7) Oxidation of the Zircaloy-steel phases takes place mainly in the reflood stage 3 of the accident scenario. (orig.) [de

  15. Characterization of fuel distribution in the Three Mile Island Unit 2 (TMI-2) reactor system by neutron and gamma-ray dosimetry

    International Nuclear Information System (INIS)

    Gold, R.; Roberts, J.H.; Ruddy, F.H.; Preston, C.C.; McNeece, J.P.; Kaiser, B.J.; McElroy, W.N.

    1984-01-01

    Neutron and gamma-ray dosimetry are being used for nondestructive assessment of the fuel distribution throughout the Three Mile Island Unit 2 (TMI-2) reactor core region and primary cooling system. The fuel content of TMI-2 makeup and purification Demineralizer A has been quantified with Si(Li) continuous gamma-ray spectrometry and solid-state track recorder (SSTR) neutron dosimetry. For fuel distribution characterization in the core region, results from SSTR neutron dosimetry exposures in the TMI-2 reactor cavity are presented. These SSTR results are consistent with the presence of a significant amount of fuel debris, equivalent to several fuel assemblies or more, lying at the bottom of the reactor vessel. (Auth.)

  16. Testing and examination of TMI-2 electrical components and discrete devices

    International Nuclear Information System (INIS)

    Soberano, F.T.

    1982-11-01

    This report discusses the approach and results of the in situ test conducted on TMI-2 reactor building electrical components and discrete devices. Also included are the necessary presumptions and assumptions to correlate observed anomalies to the accident

  17. Insights on severe accident chemistry from TMI-2

    International Nuclear Information System (INIS)

    Hobbins, R.R.; Cronenberg, A.W.; Langer, S.; Owen, D.E.; Akers, D.W.

    1986-01-01

    Chemical analyses are being carried out on materials removed from the damaged reactor at TMI-2. Characteristics of TMI-2 fuel, control, fission product and structural materials based on these analyses are presented. Emphasis is placed on chemistry within the pressure vessel although descriptions of, and postulated mechanisms for, materials transported from the vessel to the reactor building are also discussed. Indications of the oxygen potential in the reactor pressure vessel during the high temperature phase of the accident are of particular significance for the analysis of damage progression and fission product behavior during severe accidents. The results of thermodynamic and kinetic calculations for chemical species present during the high temperature portion of the accident (during core uncovery) are presented. Insights on chemistry of significance for severe accident analysis which follow from the evaluation of the TMI-2 accident are discussed. 38 refs., 9 figs., 2 tabs

  18. TMI-2 core debris analysis

    International Nuclear Information System (INIS)

    Cook, B.A.; Carlson, E.R.

    1985-01-01

    One of the ongoing examination tasks for the damaged TMI-2 reactor is analysis of samples of debris obtained from the debris bed presently at the top of the core. This paper summarizes the results reported in the TMI-2 Core Debris Grab Sample Examination and Analysis Report, which will be available early in 1986. The sampling and analysis procedures are presented, and information is provided on the key results as they relate to the present core condition, peak temperatures during the transient, temperature history, chemical interactions, and core relocation. The results are then summarized

  19. Accountability study for TMI-2 fuel

    Energy Technology Data Exchange (ETDEWEB)

    Goris, P; Scott, D D

    1981-05-01

    The TMI-2 accountability study considers problems of identifying, measuring, and accounting for TMI-2 fuel in the resident condition, as it is removed from the reactor, during subsequent cleanup, and during post-removal examinations. The goal is to identify methods and procedures which will provide a verifiable material balance equating to the pre-accident balance.

  20. Qualification of data obtained during a severe accident. Illustrative examples from TMI-2 evaluations

    Energy Technology Data Exchange (ETDEWEB)

    Rempe, Joy L. [Rempe and Associates, Idaho Falls, ID (United States); Knudson, Darrell L. [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2015-02-01

    The accidents at the Three Mile Island Unit 2 (TMI-2) Pressurized Water Reactor (PWR) and the Daiichi Units 1, 2, and 3 Boiling Water Reactors (BWRs) provide unique opportunities to evaluate instrumentation exposed to severe accident conditions. Conditions associated with the release of coolant and the hydrogen burn that occurred during the TMI-2 accident exposed instrumentation to harsh conditions, including direct radiation, radioactive contamination, and high humidity with elevated temperatures and pressures. Post-TMI-2 instrumentation evaluation programs focused on data required by TMI-2 operators to assess the condition of the reactor and containment and the effect of mitigating actions taken by these operators. Prior efforts also focused on sensors providing data required for subsequent forensic evaluations and accident simulations. This paper provides additional details related to the formal process used to develop a qualified TMI-2 data base and presents data qualification details for three parameters: reactor coolant system (RCS) pressure; containment building temperature; and containment pressure. These selected examples illustrate the types of activities completed in the TMI-2 data qualification process and the importance of such a qualification effort. These details are described to facilitate implementation of a similar process using data and examinations at the Daiichi Units 1, 2, and 3 reactors so that BWR-specific benefits can be obtained.

  1. Acquisition of TMI-2 core debris samples

    International Nuclear Information System (INIS)

    Owen, D.E.; Fee, D.S.; Bower, J.M.

    1983-01-01

    Recently the first samples of material were obtained from the damaged TMI-2 reactor core. This paper reviews the design and operation of the specialized equipment used to acquire these samples. In addition to two unique core debris rubble bed samplers, a 14-m-long sampler deployment boom, radiation shielding, and sample casks were designed and fabricated for the task. Analysis of the six TMI-2 core debris specimens obtained are presently underway at several laboratories, and the analysis plans are presented

  2. Processes, Techniques, and Successes in Welding the Dry Shielded Canisters of the TMI-2 Reactor Core Debris

    International Nuclear Information System (INIS)

    Zirker, L.R.; Rankin, R.A.; Ferrell, L.J.

    2002-01-01

    The Idaho National Engineering and Environmental Laboratory (INEEL) is operated by Bechtel-BWXT Idaho LLC (BBWI), which recently completed a very successful $100 million Three-Mile Island-2 (TMI-2) program for the Department of Energy (DOE). This complex and challenging program used an integrated multidisciplinary team approach that loaded, welded, and transported an unprecedented 25 dry shielded canisters (DSC) in seven months, and did so ahead of schedule. The program moved over 340 canisters of TMI-2 core debris that had been in wet storage into a dry storage facility at the INEEL. The main thrust of this paper is relating the innovations, techniques, approaches, and lessons learned associated to welding of the DSC's. This paper shows the synergism of elements to meet program success and shares these lessons learned that will facilitate success with welding of dry shielded canisters in other DOE complex dry storage programs

  3. Remote handling systems help TMI-2 cleanup efforts

    International Nuclear Information System (INIS)

    Taylor, G.M.

    1984-01-01

    As the cleanup at Three Mile Island-2 reactor progresses, the use of remote handling technology will play an important role in the upcoming decontamination of the reactor building basement and the defueling of the reactor. The Remote Reconnaissance Vehicle, Rover, and its use in cleanup tasks are described. Possible concepts for second-generation vehicles are discussed. Earlier less-advanced remote handling equipment used at TMI-2 are also described. Techniques planned for reactor defueling using the Remotely Operated Service Arm, Rosa, and advanced remote handling technology is presented

  4. TMI-2 quick look examination

    International Nuclear Information System (INIS)

    Franz, W.A.; Rider, R.L.; Austin, W.A.; Cole, N.

    1982-01-01

    The purpose of this work, conducted under the Department of Energy's Reactor Evaluation Program, was to gain the earliest possible access to the TMI-2 reactor vessel and to determine the condition of the plenum assembly and the reactor core. Completion of this examination has also provided substantial progress towards removal of the reactor vessel head and eventual defueling. Two methods were developed for gaining through-head access. The first involves removal of an entire CRDM, providing a 6.8-cm-diameter access through the nozzle. In case normal uncoupling proved unsuccessful, contingency techniques were developed to disconnect the leadscrew. Two contingency procedures, one ex-head and one in-head, were developed. A second technique for through-head access, the so-called Quick Look technique, was developed at the suggestion of the Technical Assessment and Advisory Group (TAAG), a group of senior technical people funded by DOE to advise GPU Nuclear on the TMI-2 recovery. This simplified method involves uncoupling and removing a CRDM leadscrew by basically normal methods and inserting a Closed Circuit Television (CCTV) camera directly through the space vacated by the leadscrew

  5. Electrometallurgical treatment of TMI-2 fuel debris

    International Nuclear Information System (INIS)

    Karell, E.J.; Gourishankar, K.V.; Johnson, G.K.

    1997-01-01

    Argonne National Laboratory (ANL) has developed an electrometallurgical treatment process suitable for conditioning DOE oxide spent fuel for long-term storage or disposal. The process consists of an initial oxide reduction step that converts the actinide oxides to a metallic form, followed by an electrochemical separation of uranium from the other fuel constituents. The final product of the process is a uniform set of stable waste forms suitable for long-term storage or disposal. The suitability of the process for treating core debris from the Three Mile Island-2 (TMI-2) reactor is being evaluated. This paper reviews the results of preliminary experimental work performed using simulated TMI-2 fuel debris

  6. Preliminary results of the TMI-2 radioactive iodine mass balance study

    International Nuclear Information System (INIS)

    Pelletier, C.A.; Cox, T.E.; Reeder, D.L.; Vollique, P.G.; Thomas, C.D.

    1982-01-01

    Analysis of samples taken from the Three Mile Island Unit 2 (TMI-2) reactor building following the 1979 accident indicates the fraction of the radioactive iodine (radioiodine) inventory in the core released to the uilding atmosphere is smaller tan assumed in Regulatory Guide 1.4. This summary presents analytical results supporting this conclusion

  7. TMI-2 health physics program

    International Nuclear Information System (INIS)

    Ruhter, P.

    1980-01-01

    The field instrumentation used in the TMI-2 decontamination is basically the Eberline standard line of instruments: the R01 Cutie Pie, the R02 and 2A boxes, and a Teletector. Personnel monitoring is done with a Harshaw TLD badge with two chips. The protective clothing is cotton PC's in multiple layers. The mix of isotopes in the TMI-2 coolant system and the health physics impact are discussed. Other potential protective clothings are discussed

  8. TMI-2 Vessel Investigation Project integration report

    International Nuclear Information System (INIS)

    Wolf, J.R.; Rempe, J.L.; Stickler, L.A.; Korth, G.E.; Diercks, D.R.; Neimark, L.A.; Akers, D.W.; Schuetz, B.K.; Shearer, T.L.; Chavez, S.A.; Thinnes, G.L.; Witt, R.J.; Corradini, M.L.; Kos, J.A.

    1994-03-01

    The Three Mile Island Unit 2 (TMI-2) Vessel Investigation Project (VIP) was an international effort that was sponsored by the Nuclear Energy Agency of the Organization for Economic Cooperation and Development. The primary objectives of the VIP were to extract and examine samples from the lower head and to evaluate the potential modes of failure and the margin of structural integrity that remained in the TMI-2 reactor vessel during the accident. This report presents a summary of the major findings and conclusions that were developed from research during the VIP. Results from the various elements of the project are integrated to form a cohesive understanding of the vessel's condition after the accident

  9. TMI-2 Vessel Investigation Project integration report

    Energy Technology Data Exchange (ETDEWEB)

    Wolf, J. R.; Rempe, J. L.; Stickler, L. A.; Korth, G. E.; Diercks, D. R.; Neimark, L. A.; Akers, D W; Schuetz, B. K.; Shearer, T L; Chavez, S. A.; Thinnes, G. L.; Witt, R. J.; Corradini, M L; Kos, J. A. [EG and G Idaho, Inc., Idaho Falls, ID (United States)

    1994-03-01

    The Three Mile Island Unit 2 (TMI-2) Vessel Investigation Project (VIP) was an international effort that was sponsored by the Nuclear Energy Agency of the Organization for Economic Cooperation and Development. The primary objectives of the VIP were to extract and examine samples from the lower head and to evaluate the potential modes of failure and the margin of structural integrity that remained in the TMI-2 reactor vessel during the accident. This report presents a summary of the major findings and conclusions that were developed from research during the VIP. Results from the various elements of the project are integrated to form a cohesive understanding of the vessel`s condition after the accident.

  10. The TMI-2 SNM accountability program

    International Nuclear Information System (INIS)

    Schork, J.S.; Rogan, R.E.; Deininger, F.W.; Weaver, W.W.

    1988-01-01

    This paper describes the Special Nuclear Material (SNM) Accountability Program for Three Mile Isaland Unit 2. The TMI-2 SNM Accountability Program is uniquely designed to inventory and control the SNM borne by the fuel materials that were distributed throughout the Reactor Vessel and connected Reactor Coolant System piping as a result of the March, 1979 accident. The current knowledge of fuel (SNM) quantities and locations as a result of the TMI-2 accident is reviewed. The inventory and control of fuel debris canisters, core debris samples, water process filters, ion exchnagers and radioactive waste that contain SNM is discussed. In addition, the methods and techniques for performing the Post-Defueling Survey of residual SNM quantities at the end of defueling activities are described. The integration of the Waste Management (shipping), Defueling (packaging), Radiological Controls and Data Management and Analysis Departments support is addressed. Finally, the contractual transfer of TMI-2 fuel debris ownership from GPU Nuclear to the Department of Energy is reviewed

  11. Surface deposition measurements of the TMI-2 gross decontamination experiment

    International Nuclear Information System (INIS)

    McIssac, C.V.; Hetzer, D.C.

    1982-01-01

    In order to measure the effectiveness of the gross decontamination experiment (principally a water spray technique) performed in the TMI-2 reactor building, the Technical Information and Examination Program's Radiation and Environment personnel made surface activity measurements before and after the experiment. In conjunction with surface sampling, thermoluminescent dosimeter (TLD) and gamma spectrometry measurements were also performed to distinguish between radiation fields and contamination. The surface sampler used to collect samples from external surfaces within the reactor building is a milling tool having four major components: a 1.27-cm constant-speed drill; a drill support assembly that allows setting sample penetration depth; filter cartridges for intake air purification and sample collection; and an air pump that forces air across the surface being sampled and through the sample filter cartridge

  12. TMI-2: Unique waste management technology

    International Nuclear Information System (INIS)

    Bixby, W.W.; Young, W.R.; Grant, P.J.

    1987-01-01

    The 1979 accident at TMI-2 severely damaged the reactor core and contaminated more than a million gallons of water. Subsequent activities created another million gallons of water. The damaged reactor core represented a new waste form and cleanup of the contaminated water and system components created other new waste forms requiring creative approaches to waste management. This paper focuses on technologies that were developed specific to fuel waste management, core debris shipping, processing accident generated water, and disposal of the resultant waste forms

  13. Lessons learned from hydrogen generation and burning during the TMI-2 event

    International Nuclear Information System (INIS)

    Henrie, J.O.; Postma, A.K.

    1987-05-01

    This document summarizes what has been learned from generation of hydrogen in the reactor core and the hydrogen burn that occurred in the containment building of the Three Mile Island Unit No. 2 (TMI-2) nuclear power plant on March 28, 1979. During the TMI-2 loss-of-coolant accident (LOCA), a large quantity of hydrogen was generated by a zirconium-water reaction. The hydrogen burn that occurred 9 h and 50 min after the initiation of the TMI-2 accident went essentially unnoticed for the first few days. Even through the burn increased the containment gas temperature and pressure to 1200 0 F (650 0 C) and 29 lb/in 2 (200 kPa) gage, there was no serious threat to the containment building. The processes, rates, and quantities of hydrogen gas generated and removed during and following the LOCA are described in this report. In addition, the methods which were used to define the conditions that existed in the containment building before, during, and after the hydrogen burn are described. The results of data evaluations and engineering calculations are presented to show the pressure and temperature histories of the atmosphere in various containment segments during and after the burn. Material and equipment in reactor containment buildings can be protected from burn damage by the use of relatively simple enclosures or insulation

  14. Reactor building

    International Nuclear Information System (INIS)

    Ebata, Sakae.

    1990-01-01

    At least one valve rack is disposed in a reactor building, on which pipeways to a main closure valve, valves and bypasses of turbines are placed and contained. The valve rack is fixed to the main body of the building or to a base mat. Since the reactor building is designed as class A earthquake-proofness and for maintaining the S 1 function, the valve rack can be fixed to the building main body or to the base mat. With such a constitution, the portions for maintaining the S 1 function are concentrated to the reactor building. As a result, the dispersion of structures of earthquake-proof portion corresponding to the reference earthquake vibration S 1 can be prevented. Accordingly, the conditions for the earthquake-proof design of the turbine building and the turbine/electric generator supporting rack are defined as only the class B earthquake-proof design conditions. In view of the above, the amount of building materials can be saved and the time for construction can be shortened. (I.S.)

  15. TMI-2 spent fuel shipping

    International Nuclear Information System (INIS)

    Quinn, G.J.; Burton, H.M.

    1985-01-01

    TMI-2 failed fuel will be shipped to the Idaho National Engineering Laboratory for use in the DOE Core Examination Program. The fuel debris will be loaded into three types of canisters during defueling and dry loaded into a spent fuel shipping cask. The cask design accommodates seven canisters per cask and has two separate containment vessels with ''leaktight'' seals. Shipments are expectd to begin in early 1986

  16. TMI-2 isotopic inventory calculations

    Energy Technology Data Exchange (ETDEWEB)

    Schnitzler, B G; Briggs, J B

    1985-08-01

    Point isotopic depletion methods are used to develop spatially dependent fission product and heavy metal inventories for the TMI-2 core. Burnup data from 1239 fuel nodes (177 elements, 7 axial nodes per element) are utilized to preserve the core axial and radial power distributions. A full-core inventory is calculated utilizing 12 fuel groups (four burnup ranges for each of three initial enrichments). Calculated isotopic ratios are also presented as a function of burnup for selected nuclides. Specific applications of the isotopic ratio data include correlation of fuel debris samples with core location and estimates of fission product release fractions. 24 figs., 25 tabs.

  17. TMI-2 data qualification and data bases

    International Nuclear Information System (INIS)

    Golden, D.W.; Anderson, J.L.; Brower, R.W.; Fackrell, L.J.; McCormick, R.D.

    1988-01-01

    TMI-2 data are of great interest to the scientific community, since the March, 1979 accident represents the only occurrence of a transient resulting in severe fuel damage in a full-scale commercial production reactor. This paper discusses the data sources available for accident analyses. Techniques to qualify and quantify time series measurement and sample data which have been collected and evaluated to this time are discussed. Data base products are described which have been developed for use by analysts investigating the many aspects of the accident. Applications processing features have been incorporated into the software designs to provide the end user with many useful capabilities in sorting, viewing and analyzing the data; data base discussions emphasize these features. 14 refs., 5 figs

  18. Sampling and examination methods used for TMI-2 samples

    International Nuclear Information System (INIS)

    Marley, A.W.; Akers, D.W.; McIsaac, C.V.

    1988-01-01

    The purpose of this paper is to summarize the sampling and examination techniques that were used in the collection and analysis of TMI-2 samples. Samples ranging from auxiliary building air to core debris were collected and analyzed. Handling of the larger samples and many of the smaller samples had to be done remotely and many standard laboratory analytical techniques were modified to accommodate the extremely high radiation fields associated with these samples. The TMI-2 samples presented unique problems with sampling and the laboratory analysis of prior molten fuel debris. 14 refs., 8 figs

  19. Qualification of Daiichi Units 1, 2, and 3 Data for Severe Accident Evaluations - Process and Illustrative Examples from Prior TMI-2 Evaluations

    Energy Technology Data Exchange (ETDEWEB)

    Rempe, Joy Lynn [Idaho National Lab. (INL), Idaho Falls, ID (United States); Knudson, Darrell Lee [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2014-09-01

    The accidents at the Three Mile Island Unit 2 (TMI-2) Pressurized Water Reactor (PWR) and the Daiichi Units 1, 2, and 3 Boiling Water Reactors (BWRs) provide unique opportunities to evaluate instrumentation exposed to severe accident conditions. Conditions associated with the release of coolant and the hydrogen burn that occurred during the TMI-2 accident exposed instrumentation to harsh conditions, including direct radiation, radioactive contamination, and high humidity with elevated temperatures and pressures. As part of a program initiated in 2012 by the Department of Energy Office of Nuclear Energy (DOE-NE), a review was completed to gain insights from prior TMI-2 sensor survivability and data qualification efforts. This initial review focused on the set of sensors deemed most important by post-TMI-2 instrumentation evaluation programs. Instrumentation evaluation programs focused on data required by TMI-2 operators to assess the condition of the reactor and containment and the effect of mitigating actions taken by these operators. In addition, prior efforts focused on sensors providing data required for subsequent forensic evaluations and accident simulations. To encourage the potential for similar activities to be completed for qualifying data from Daiichi Units 1, 2, and 3, this report provides additional details related to the formal process used to develop a qualified TMI-2 data base and presents data qualification details for three parameters: primary system pressure; containment building temperature; and containment pressure. As described within this report, sensor evaluations and data qualification required implementation of various processes, including comparisons with data from other sensors, analytical calculations, laboratory testing, and comparisons with sensors subjected to similar conditions in large-scale integral tests and with sensors that were similar in design to instruments easily removed from the TMI-2 plant for evaluations. As documented

  20. Calibration of SSTR neutron dosimetry for TMI-2 applications

    International Nuclear Information System (INIS)

    Gold, R.; Ruddy, F.H.; Roberts, J.H.; Preston, C.C.; Ulseth, J.A.; McElroy, W.N.; Leitz, F.J.; Hayward, B.R.; Schmittroth, F.A.

    1982-01-01

    Application of neutron dosimetry for assessment of fuel distribution throughout the Three Mile Island-2 (TMI-2) reactor-core region and the primary-coolant system is advanced. Neutron dosimetry in the reactor cavity, i.e. the cavity between the pressure vessel and the biological shield, could provide data for the assessment of the core fuel distribution. A more immediate task entails locating and quantifying the amount of fuel debris in the ex-core primary coolant system; in the range of 1 to 1000 kg. Solid-state track-recorder (SSTR) neutron dosimetry is considered for such exploratory scoping experiments at TMI-2. The sensitivity of mica- 235 U (asymptotically thick) SSTR has been ascertained for such environments. It has been demonstrated that the SSTR method has adequate sensitivity to properly respond and detect fuel quantities of the order of 1 kg in the ex-core primary coolant system. 21 figures

  1. Criticality safety implementation, operations, and training for TMI-2 defueling

    International Nuclear Information System (INIS)

    Knief, R.A.; Fergus, I.E.

    1986-01-01

    The nuclear criticality safety of the accident-damaged Three Mile Island Unit 2 (TMI-2) reactor has depended primarily on the use of soluble neutron poison. Detailed calculations have been performed to establish the boric acid concentrations needed for subcriticality of postulated core configurations related to recovery and defueling activities. Based on the minimum acceptable boron concentration, operating requirements were developed, safety reviews conducted, strategies and procedures implemented, and training conducted

  2. March 28, 1979 plus 42 months, or a status report on the TMI-2 cleanup program

    International Nuclear Information System (INIS)

    Dieckamp, Herman

    1982-01-01

    The author gives a general overview of the TMI-2 cleanup program, including regulatory, financial, and public or political aspects. He reviews some of the major technical accomplishments, including the controlled venting of krypton, processing of more than half a million gallons of water from the auxiliary building using the EPICOR-II process, decontamination of over 600,000 gallons of water from the containment building using the submerged demineralizer system, processing of water from the reactor coolant system, and manned entries into the containment building. Cleaning up after a major reactor accident is expensive, not only because of the hostile physical environment in which the work must be carried out, but also because of regulatory, political and public acceptance constraints. The technological ingredients of a cleanup program exist, but the task of assembling those ingredients, ensuring a balance between cost, schedule and risk, and selling the resulting program to regulators and the public is a demanding one

  3. TMI-2 as a major experiment

    International Nuclear Information System (INIS)

    Kinter, E.E.

    1994-01-01

    This paper shows how TMI-2 accident changed profoundly the nuclear research perspective with a large-scale core damaging accident which provided an opportunity to compare theory and computer codes to an enormous set of benchmark data. The vessel sample gathering and study programs were unusually successful. The TMI-2 events appeared to dramatically challenge not only the engineering designs and safety calculations, but also the institutional arrangements between industry, state, ..., causing the establishment of the INPO, changes in management relationships, etc

  4. Proceedings of the US Nuclear Regulatory Commission fifteenth water reactor safety information meeting: Volume 6, Decontamination and decommissioning, accident management, TMI-2

    Energy Technology Data Exchange (ETDEWEB)

    Weiss, A. J. [comp.

    1988-02-01

    This six-volume report contains 140 papers out of the 164 that were presented at the Fifteenth Water Reactor Safety Information Meeting held at the National Bureau of Standards, Gaithersburg, Maryland, during the week of October 26-29, 1987. The papers are printed in the order of their presentation in each session and describe progress and results of programs in nuclear safety research conducted in this country and abroad. This report, Volume 6, discusses decontamination and decommissioning, accident management, and the Three Mile Island-2 reactor accident. Thirteen reports have been cataloged separately.

  5. Proceedings of the US Nuclear Regulatory Commission fifteenth water reactor safety information meeting: Volume 6, Decontamination and decommissioning, accident management, TMI-2

    International Nuclear Information System (INIS)

    Weiss, A.J.

    1988-02-01

    This six-volume report contains 140 papers out of the 164 that were presented at the Fifteenth Water Reactor Safety Information Meeting held at the National Bureau of Standards, Gaithersburg, Maryland, during the week of October 26-29, 1987. The papers are printed in the order of their presentation in each session and describe progress and results of programs in nuclear safety research conducted in this country and abroad. This report, Volume 6, discusses decontamination and decommissioning, accident management, and the Three Mile Island-2 reactor accident. Thirteen reports have been cataloged separately

  6. MELCOR analysis of the TMI-2 accident

    International Nuclear Information System (INIS)

    Boucheron, E.A.

    1990-01-01

    This paper describes the analysis of the Three Mile Island-2 (TMI-2) standard problem that was performed with MELCOR. The MELCOR computer code is being developed by Sandia National Laboratories for the Nuclear Regulatory Commission for the purpose of analyzing severe accident in nuclear power plants. The primary role of MELCOR is to provide realistic predictions of severe accident phenomena and the radiological source team. The analysis of the TMI-2 standard problem allowed for comparison of the model predictions in MELCOR to plant data and to the results of more mechanistic analyses. This exercise was, therefore valuable for verifying and assessing the models in the code. The major trends in the TMI-2 accident are reasonably well predicted with MELCOR, even with its simplified modeling. Comparison of the calculated and measured results is presented and, based on this comparison, conclusions can be drawn concerning the applicability of MELCOR to severe accident analysis. 5 refs., 10 figs., 3 tabs

  7. TMI-2 Core Shipping Preparations

    International Nuclear Information System (INIS)

    Ball, L.J.; Barkanic, R.J.; Conaway, W.T. II; Schmoker, D. S.; Post, Roy G.

    1988-01-01

    Shipping the damaged core from the Unit 2 reactor of Three Mile Island Nuclear Power Station near Harrisburg, PA, to the Idaho National Engineering Laboratory near Idaho Falls, ID, required development and implementation of a completely new spent fuel transportation system. This paper describes of the equipment developed, the planning and activities used to implement the hard-ware-systems into the facilities, and the planning involved in making the rail shipments. It also includes a summary of recommendations resulting from this experience. (author)

  8. TMI-2 information and examination program 1981 annual report

    International Nuclear Information System (INIS)

    1982-04-01

    The Department of Energy's Technical Information and Examination Program at Three Mile Island Unit 2 continued the research and development work begun on the Island in 1979. The work concentrated in seven major areas: instrumentation and electrical components; radiation and environment; off-site core examination; radioactive waste technology development; configuration and document control; waste immobilization; and reactor evaluation. Research and development work associated with the program aims toward communicating applicable information to the nuclear community. The program seeks to assist in resolving specific problems at TMI-2 and to stimulate interest in specific work activities, thus ensuring that the entire nuclear industry avails itself of the maximum amount of information possible

  9. The TMI-2 core relocation: Heat transfer and mechanism

    International Nuclear Information System (INIS)

    Epstein, M.; Fauske, H.K.

    1987-07-01

    It is postulated that the collapse of the upper debris bed was the main cause of core failure and core material relocation during the TMI-2 accident. It is shown that this mechanism of core relocation can account for the timescale(s) and energy transfer rate inferred from plant instrumentation. Additional analysis suggests that the water in the lower half of the reactor vessel was subcooled at the onset of relocation, as subcooling serves to explain the final coolable configuration at the bottom of the TMI vessel

  10. TMI-2 core debris-cesium release/settling test. Volume 3

    Energy Technology Data Exchange (ETDEWEB)

    Akers, D W; Johnson, D A

    1984-12-01

    Cesium release, turbidity and airborne potential tests were conducted on 50 grams of TMI-2 core debris materials. The tests were performed on the debris in two conditions: on the as-received core debris specimen, and after crushing the debris to alter the particle size distribution. The crushing was intended to simulate the breakup of TMI-2 core material that may occur during reactor defueling. These tests are intended to assist GPU Nuclear in predicting the effect of defueling on the reactor environment.

  11. Summary of TMI-2 data bases

    International Nuclear Information System (INIS)

    Brower, R.W.

    1987-09-01

    This report summarizes seven major data base products produced by the Data Reduction and qualification Section of the TMI-2 Accident Evaluation Program. The purpose and a brief description of data base structure are presented in the introductory section, together with rational involved in selection of data base media. Major emphasis in the report is placed in more detailed examinations of four personal computer data bases which utilize an INEL developed data base management system, SAGE. Content of each data base is described, the current development status is defined and future activity associated with each dynamic structures is outlined

  12. TMI-2 fuel canister and core sample handling equipment used in INEL hot cells

    International Nuclear Information System (INIS)

    McConnell, J.W. Jr.; Shurtliff, W.T.; Lynch, R.J.; Croft, K.M.; Whitmill, L.J.; Allen, S.M.

    1987-01-01

    This paper describes the specialized remote handling equipment developed and used at the Idaho National Engineering Laboratory (INEL) to handle samples obtained from the core of the damaged Unit 2 reactor at Three Mile Island Nuclear Power Station (TMI-2). Samples of the core were removed, placed in TMI-2 fuel canisters, and transported to the INEL. Those samples will be examined as part of the analysis of the TMI-2 accident. The equipment described herein was designed for removing sample materials from the fuel canisters, assisting with initial examinations, and processing samples in preparation for detailed examinations. The more complex equipment used microprocessor remote controls with electric motor drives providing the required force and motion capabilities. The remaining components were unpowered and manipulator assisted

  13. TMI-2 Technical Information and Examination Program. 1982 annual report

    International Nuclear Information System (INIS)

    1983-04-01

    The Department of Energy's Technical Information and Examination Program at Three Mile Island Unit 2 continued the research and development work begun on the island in 1980. The work concentrated in seven major areas: instrumentation and electrical components, radiation and environment, core activities, information and industry coordination, configuration and document control, waste immobilization, and reactor evaluation. The program assists in resolving specific problems at TMI-2 while developing techniques and broadening understanding of accident consequences to improve the overall safety and reliability of nuclear power. The Technical Information and Examination Program aims to communicate applicable information to the nuclear power industry to ensure that the industry can avail itself of the maximum amount of information possible

  14. TMI-2 Technical Information and Examination Program 1983 annual report

    International Nuclear Information System (INIS)

    Scardena, D.E.

    1984-04-01

    The Department of Energy's Technical Information and Examination Program at Three Mile Island Unit 2 continued the research and development work begun on the Island in 1980. The work concentrated in six major areas: waste immobilization, reactor evaluation, data acquisition, information and industry coordination, core activities, and EPICOR II and waste research and disposition. The program assists in resolving specific problems at TMI-2 while developing techniques and broadening understanding of accident consequences to improve the overall safety and reliability of nuclear power. The Technical Information and Examination Program aims to communicate applicable information to the nuclear power industry to ensure that the industry can avail itself to the maximum amount of information possible

  15. TMI-2 [Three Mile Island Unit 2] licensing history

    International Nuclear Information System (INIS)

    Byrne, J.J.

    1988-01-01

    Three Mile Island (TMI), which is located in central Pennsylvania near Harrisburg, is the site of the TMI-2 accident, the most significant nuclear accident in US commercial nuclear power. Since the accident on March 28, 1979, TMI-2 has been undergoing cleanup activities designed to place the plant in a safe, stable, and secure postaccident configuration. At the completion of the cleanup program, TMI-2 will be placed in such a configuration, termed postdefueling monitored storage (PDMS), by the licensee, GPU Nuclear Corporation. The purpose of this paper is to provide a brief overview of the TMI-2 licensing history and to describe its impact on the regulatory process

  16. TMI-2 - A Case Study for PWR Instrumentation Performance during a Severe Accident

    Energy Technology Data Exchange (ETDEWEB)

    Joy L. Rempe; Darrell L. Knudson

    2014-05-01

    The accident at the Three Mile Island Unit 2 (TMI-2) reactor provided a unique opportunity to evaluate sensors exposed to severe accident conditions. Conditions associated with the release of coolant and the hydrogen burn that occurred during this accident exposed instrumentation to harsh conditions, including direct radiation, radioactive contamination, and high humidity with elevated temperatures and pressures. As part of a program initiated in 2012 by the Department of Energy Office of Nuclear Energy (DOE-NE), a review was completed to gain insights from prior TMI-2 sensor survivability and data qualification efforts. This new effort focussed upon a set of sensors that provided critical data to TMI-2 operators for assessing the condition of the plant and the effects of mitigating actions taken by these operators. In addition, the effort considered sensors providing data required for subsequent accident simulations. Over 100 references related to instrumentation performance and post-accident evaluations of TMI-2 sensors and measurements were reviewed. Insights gained from this review are summarized within this report. For each sensor, a description is provided with the measured data and conclusions related to the sensor’s survivability, and the basis for conclusions about its survivability. As noted within this document, several techniques were invoked in the TMI-2 post-accident evaluation program to assess sensor status, including comparisons with data from other sensors, analytical calculations, laboratory testing, and comparisons with sensors subjected to similar conditions in large-scale integral tests and with sensors that were similar in design but more easily removed from the TMI-2 plant for evaluations. Conclusions from this review provide important insights related to sensor survivability and enhancement options for improving sensor performance. In addition, this document provides recommendations related to the sensor survivability and data evaluation

  17. TMI-2 - A Case Study for PWR Instrumentation Performance during a Severe Accident

    Energy Technology Data Exchange (ETDEWEB)

    Joy L. Rempe; Darrell L. Knudson

    2013-03-01

    The accident at the Three Mile Island Unit 2 (TMI-2) reactor provided a unique opportunity to evaluate sensors exposed to severe accident conditions. Conditions associated with the release of coolant and the hydrogen burn that occurred during this accident exposed instrumentation to harsh conditions, including direct radiation, radioactive contamination, and high humidity with elevated temperatures and pressures. As part of a program initiated in 2012 by the Department of Energy Office of Nuclear Energy (DOE-NE), a review was completed to gain insights from prior TMI-2 sensor survivability and data qualification efforts. This new effort focussed upon a set of sensors that provided critical data to TMI-2 operators for assessing the condition of the plant and the effects of mitigating actions taken by these operators. In addition, the effort considered sensors providing data required for subsequent accident simulations. Over 100 references related to instrumentation performance and post-accident evaluations of TMI-2 sensors and measurements were reviewed. Insights gained from this review are summarized within this report. For each sensor, a description is provided with the measured data and conclusions related to the sensor’s survivability, and the basis for conclusions about its survivability. As noted within this document, several techniques were invoked in the TMI-2 post-accident evaluation program to assess sensor status, including comparisons with data from other sensors, analytical calculations, laboratory testing, and comparisons with sensors subjected to similar conditions in large-scale integral tests and with sensors that were similar in design but more easily removed from the TMI-2 plant for evaluations. Conclusions from this review provide important insights related to sensor survivability and enhancement options for improving sensor performance. In addition, this document provides recommendations related to the sensor survivability and data evaluation

  18. Present status of TMI-2 plant and results of its research

    International Nuclear Information System (INIS)

    Sasaki, Sadaaki; Yokomi, Michiro.

    1987-01-01

    In the accident occurred in the TMI-2 plant on March 28, 1979, the damage was caused in the reactor core, but there scarcely was the effect on the health and safety of general public around the power station. But in USA, it was decided to collect the data on the fuel, decontamination, waste management and so on of this plant and to advance the survey and research on the safety by the analysis and evaluation of the course of the accident mainly by GPUN, EPRI, NRC and DOE. Also in Japan, it was judged that the participation in this research would be useful for improving the reliability of Japanese nuclear power plants hereafter, and the Japan-USA agreement on TMI-2 research and development project was concluded on April 16, 1984. The activity plan in TMI-2 is divided into three stages. Phase 1 is the stage of stabilization, Phase 2 is the stage of taking fuel out, and Phase 3 is the stage of cleaning. At present, Phase 2 - 3 are in progress, and the taking-out and transport of fuel and decontamination are carried out. After finishing Phase 3, the TMI-2 plant is placed in the state of monitoring and preservation, which is scheduled in September, 1988. The final disposal of the plant will be determined thereafter. Decontamination, treatment of contaminated water and wastes, taking-out and transport of fuel, state of the reactor core and others are reported. (Kako, I.)

  19. Technology transfer and radioactive waste management at TMI-2 [Three Mile Island Unit 2

    International Nuclear Information System (INIS)

    Saunders, J.R.

    1988-01-01

    The accident that occurred on March 28, 1979, at the Three Mile Island Unit 2 (TMI-2) nuclear generating station caused extensive damage to the reactor core and created high radiation contamination levels throughout the facility. The electric power industry, regulators, and government agencies were faced with one of the most technically challenging recovery situations ever encountered in this country. But it was also realized that this adversity presented opportunities for the advancement of state-of-the-art technologies as well as the potential to produce information that could enhance nuclear power plant safety and reliability. Perhaps one of the more significant aspects of the TMI-2 recovery has been the advancement of radioactive waste management technology. The high levels and unusual nature of the TMI-2 radioactive waste necessitated the development of innovative techniques for processing, packaging, shipping, and disposal. The investment in research was rewarded with large volume reductions and associated cost savings. It is anticipated that the TMI-2 radioactive waste management technology will make major contributions to the design of new systems to meet this growing need. The following areas appear particularly suited for this purpose: volume reduction, high-integrity containers, and selective isotope removal

  20. TMI-2 core-examination program: INEL facilities readiness study

    International Nuclear Information System (INIS)

    McLaughlin, T.B.

    1983-02-01

    This report reviews the capability and readiness of remote handling facilities at the Idaho National Engineering Laboratory (INEL) to receive, and store the TMI-2 core, and to examine and analyze TMI-2 core samples. To accomplish these objectives, the facilities must be able to receive commercial casks, unload canisters from the casks, store the canisters, open the canisters, handle the fuel debris and assemblies, and perform various examinations. The report identifies documentation, including core information, necessary to INEL before receiving the entire TMI-2 core. Also identified are prerequisites to INEL's receipt of the first canister: costs, schedules, and a preliminary project plan for the tasks

  1. Investigation of hydrogen-burn damage in the Three Mile Island Unit 2 reactor building

    International Nuclear Information System (INIS)

    Alvares, N.J.; Beason, D.G.; Eidem, G.R.

    1982-06-01

    About 10 hours after the March 28, 1979 Loss-of-Coolant Accident began at Three Mile Island Unit 2, a hydrogen deflagration of undetermined extent occurred inside the reactor building. Examinations of photographic evidence, available from the first fifteen entries into the reactor building, yielded preliminary data on the possible extent and range of hydrogen burn damage. These data, although sparse, contributed to development of a possible damage path and to an estimate of the extent of damage to susceptible reactor building items. Further information gathered from analysis of additional photographs and samples can provide the means for estimating hydrogen source and production rate data crucial to developing a complete understanding of the TMI-2 hydrogen deflagration. 34 figures

  2. TMI-2 remote cameras, teleoperators, and other potential nuclear applications

    International Nuclear Information System (INIS)

    Pavelek, M.D. II

    1987-01-01

    Evaluation, testing, and use of fixed-position, remotely operated video-camera equipment and various teleoperators at Three Mile Island Nuclear Station Unit 2 (TMI-2) have demonstrated that significant potential advantages may be obtained by installing remotely operated equipment at other nuclear facilities. Teleoperators at TMI-2 have permitted successful operations in areas not accessible to personnel. Insights gained from the remotely operated camera systems and teleoperators used and evaluated at TMI-2 are discussed from the perspective of actual applications and the potential benefits from deployment of remote devices at other nuclear facilities. Actual exposure savings achieved are not presented because these applications and resultant exposure savings are specific to TMI-2. 3 refs., 4 figs

  3. Nuclear reactor building

    Energy Technology Data Exchange (ETDEWEB)

    Oshima, Nobuaki.

    1991-08-09

    The secondary container in a nuclear reactor building is made of a transparent structure having a shielding performance such as lead glass, by which the inside of the secondary container can be seen without undergoing radiation exposure. In addition, an operator transportation facility capable of carrying about 5 to 10 operators at one time is disposed, and the side of the facility on the secondary container is constituted with a transparent material such as glass, to provide a structure capable of observing the inside of the secondary container. The ventilation and air conditioning in the operator's transportation facility is in communication with the atmosphere of a not-controlled area. Accordingly, operators at the outside of the reactor building can reach the operator's transportation facility without taking and procedures for entering the controlled area and without undergoing radiation exposure. The inside of the secondary container in the reactor building can be seen from various directions through the transparent structure having the shielding performance. (N.H.).

  4. Task plan for the US Department of Energy TMI-2 programs

    International Nuclear Information System (INIS)

    1982-10-01

    The Task Plan for the US Department of Energy (DOE) Three Mile Island (TMI) Unit 2 Programs identifies the tasks to be planned and administered by the DOE Technical Integration Office (TIO) in a manner which will maximize the use of available resources, obtain the maximum benefit from the opportunities associated with the TMI-2 cleanup effort, and retrieve generically useful information for addressing some of the key problems and issues facing the nuclear power industry. The Plan identifies tasks in three major program areas where DOE has assumed implementation responsibility. The DOE TMI-2 Programs are: Data Acquisition Program, Waste Immobilization Program, and Reactor Evaluation Program. The plan is intended to serve as a management overview by defining the task objective, benefits, and work scope with respect to prioritization of tasks and utilization of resources

  5. TMI-2 transportation program - Design considerations for the NUPAC 125-B cask handling and loading/unloading equipment

    International Nuclear Information System (INIS)

    Schmoker, D.S.; Schmitt, R.C.; Barkanic, R.J.

    1987-01-01

    Removal, transport and receipt of core debris from the damaged reactor at Three Mile Island Unit 2 (TMI-2) required the design of transport cask handling and ''dry'' loading and unloading equipment. The system for ''dry'' (not underwater) loading of the transport cask includes; 1) equipment for handling the cask, 2) equipment for loading core debris canisters into the cask at TMI, and 3) equipment for removing the canisters in a hot cell facility. This paper reviews the technical design considerations operational parameters and summarizes lessons learned in the design, testing, startup and use of the equipment provided for the TMI-2 Transportation Program

  6. Simulation of a TMI-2 type scenario at the ROSA-IV program's large scale test facility: a first look

    International Nuclear Information System (INIS)

    Schultz, R.R.; Kukita, Yutaka; Tasaka, Kanji

    1984-09-01

    The Three Mile Island-2 (TMI-2) type scenario in a Westinghouse (W) type four loop pressurized water reactor (PWR) was studied in preparation for upcoming tests in the ROSA-IV Program's Large Scale Test Facility (LSTF). The LSTF is a 1/48 scale simulator of a W type four loop PWR with full scale component elevation differences. TMI-2 scenario simulation analyses were conducted to establish a pretest prediction data base for RELAP5 code evaluation purposes and to furnish a means of evaluating the LSTF's capability to simulate the reference PWR. The basis for such RELAP5 calculations and the similarities between the LSTF and reference PWR thermal-hydraulic behaviors during a TMI-2 type scenario are presented. (author)

  7. Historical summary of the fuel and waste handling and disposition activities of the TMI-2 Information and Examination Program (1980-1988)

    International Nuclear Information System (INIS)

    Reno, H.W.; Schmitt, R.C.

    1988-10-01

    This report is a historical summary of the major activities conducted by the TMI-2 Information and Examination Program in managing fuel and special radioactive wastes resulting from the accident at the Unit 2 reactor of the Three Mile Island Nuclear Power Station (TMI-2). The activities often required the development and use of advanced handling, processing, and/or disposal technologies for those wastes

  8. TMI-2 criticality studies: lower-vessel rubble and analytical benchmarking

    International Nuclear Information System (INIS)

    Westfall, R.M.; Knight, J.R.; Fox, P.B.; Herman, O.W.; Turner, J.C.

    1985-12-01

    A bounding strategy has been adopted for assuring subcriticality during all TMI-2 defueling operations. The strategy is based upon establishing a safe soluble boron level for the entire reactor core in an optimum reactivity configuration. This paper presents the determination of a fuel rubble model which yields a maximum infinite lattice multiplication factor and the subsequent application of cell-averaged constants in finite system analyses. Included in the analyses are the effects of fuel burnup determined from a simplified power history of the reactor. A discussion of the analytical methods employed and the determination of an analytical bias with benchmark crictical experiments completes the presentation. 17 tabs

  9. Overview of the TMI-2 core-examination plan

    International Nuclear Information System (INIS)

    Sumpter, K.C.; Trickeit, K.A.; Feinauer, E.; Owen, D.E.; Martin, M.R.

    1982-01-01

    The TMI-2 Core Examination Plan presents a logical organization for the sampling, examination, and ultimate utilization of the data made available from the recovery program of the TMI-2 Nuclear Plant. The plan emphasizes cooperative efforts not only between the defueling team and those dedicated to the analysis of the data but also among the various laboratories and commercial facilities participating in the program. Dealing with the defueling sequence the core examination addresses three basic objectives; understanding severe core damage initiation, propagation, and termination; supporting the technical basis for existing regulation; and improving LWR design and operation. The mere understanding of the March 1979 accident mandate fundamental reasons for examining the TMI-2 core. During all phases of the recovery effort the plan intends to utilize the information used to assist the actual defueling operation

  10. Evaluation of nuclear facility decommissioning projects. Three Mile Island Unit 2 reactor building decontamination. Summary status report. Volume 2

    International Nuclear Information System (INIS)

    Doerge, D.H.; Miller, R.L.; Scotti, K.S.

    1986-05-01

    This document summarizes information relating to decontamination of the Three Mile Island Unit 2 (TMI-2) reactor building. The report covers activities for the period of June 1, 1979 through March 29, 1985. The data collected from activity reports, reactor containment entry records, and other sources were entered into a computerized data system which permits extraction/manipulation of specific information which can be used in planning for recovery from an accident similar to that experienced at TMI-2 on March 28, 1979. This report contains summaries of man-hours, manpower, and radiation exposures incurred during decontamination of the reactor building. Support activities conducted outside of radiation areas are excluded from the scope of this report. Computerized reports included in this document are: a chronological summary listing work performed relating to reactor building decontamination for the period specified; and summary reports for each major task during the period. Each task summary is listed in chronological order for zone entry and subtotaled for the number of personnel entries, exposures, and man-hours. Manually-assembled table summaries are included for: labor and exposures by department and labor and exposures by major activity

  11. TMI-2 fuel-recovery plant. Feasibility study

    International Nuclear Information System (INIS)

    Evans, D.L.

    1982-12-01

    This project is a feasibility study for constructing a TMI-2 core Fuel Recovery Plant at the Idaho National Engineering Laboratory (INEL). The primary objectives of the Fuel Recovery Plant (FRP) are to recover and account for the fuel and to process, isolate, and package the waste material from the TMI-2 core. This feasibility study is predicated on a baseline plant and covers its design, fabrication, installation, testing and operation. Alternative methods for the disposal of the TMI-2 core have also been considered, but not examined in detail for their feasibility. The FRP will receive TMI-2 fuel in canisters. The fuel will vary from core debris to intact fuel assemblies and include some core structural materials. The canister contents will be shredded and subsequently fed to a dissolver. Uranium, plutonium, fission products, and some core structural material will be dissolved. The uranium will be separated by solvent extraction and solidified by calcination. The plutonium will also be separated by solvent extraction and routed to the Plutonium Extraction Facility. The wastes will be packaged for further treatment, temporary storage or permanent disposal

  12. TMI-2 Lessons Learned Task Force. Final report

    International Nuclear Information System (INIS)

    1979-10-01

    In its final report reviewing the Three Mile Island accident, the TMI-2 Lessons Learned Task Force has suggested change in several fundamental aspects of basic safety policy for nuclear power plants. Changes in nuclear power plant design and operations and in the regulatory process are discussed in terms of general goals. The appendix sets forth specific recommendations for reaching these goals

  13. Recommended HPI [High Pressure Injection] rates for the TMI-2 analysis exercise (0 to 300 minutes)

    International Nuclear Information System (INIS)

    Anderson, J.L.

    1987-09-01

    An international analysis exercise has been organized to evaluate the ability of nuclear reactor severe accident computer codes to predict the TMI-2 accident sequence and core damage progression during the first 300 minutes of the accident. A required boundary condition for the analysis exercise is the High Pressure Injection or make-up rates into the primary system during the accident. Recommended injection rates for the first 300 minutes of the accident are presented. Recommendations for several sensitivity studies are also presented. 6 refs., 5 figs., 1 tab

  14. TMI-2 instrument nozzle examinations at Argonne National Laboratory, February 1991--June 1993

    Energy Technology Data Exchange (ETDEWEB)

    Neimark, L.A.; Shearer, T.L.; Purohit, A.; Hins, A.G.

    1994-06-01

    The accident at the Three Mile Island Unit 2 (TMI-2) reactor in March 1979 resulted in the relocation of approximately 19,000 kg of molten core material to the lower head of the reactor vessel. This material caused extensive damage to the instrument guide tubes and nozzles and was suspected of having caused significant metallurgical changes in the condition of the lower head itself. These changes and their effect on the margin-to-failure of the lower head became the focal point of an investigation co-sponsored by the United States Nuclear Regulatory Commission (NRC) and the Organization for Economic Co-operation and Development (OECD). The TMI-2 Vessel Investigation Project (VIP) was formed to determine the metallurgical state of the vessel at the lower head and to assess the margin-to-failure of the vessel under the conditions existing during the accident. This report was prepared under the auspices of the OECD/NEA Three Mile Island Vessel Investigation Project. Under the auspices of the VIP, specimens of the reactor vessel were removed in February 1990 by MPR Associates, Inc. In addition to these specimens, fourteen instrument nozzle segments and two segments of instrument guide tubes were retrieved for metallurgical evaluation. The purpose of this evaluation was to provide additional information on the thermal conditions on the lower head that would influence the margin-to-failure, and to provide insight into the progression of the accident scenario, specifically the movement of the molten fuel across the lower head.

  15. Analysis of the TMI-2 source range detector response

    International Nuclear Information System (INIS)

    Carew, J.F.; Diamond, D.J.; Eridon, J.M.

    1980-01-01

    In the first few hours following the TMI-2 accident large variations (factors of 10-100) in the source range (SR) detector response were observed. The purpose of this analysis was to quantify the various effects which could contribute to these large variations. The effects evaluated included the transmission of neutrons and photons from the core to detector and the reduction in the multiplication of the Am-Be startup sources, and subsequent reduction in SR detector response, due to core voiding. A one-dimensional ANISN slab model of the TMI-2 core, core externals, pressure vessel and containment has been constructed for calculation of the SR detector response and is presented

  16. TMI-2 core debris analytical methods and results

    International Nuclear Information System (INIS)

    Akers, D.W.; Cook, B.A.

    1984-01-01

    A series of six grab samples was taken from the debris bed of the TMI-2 core in early September 1983. Five of these samples were sent to the Idaho National Engineering Laboratory for analysis. Presented is the analysis strategy for the samples and some of the data obtained from the early stages of examination of the samples (i.e., particle size-analysis, gamma spectrometry results, and fissile/fertile material analysis)

  17. Quality assurance in the removal and transport of the TMI-2 [Three Mile Island Unit 2] core

    International Nuclear Information System (INIS)

    Hayes, G.R.; Marsden, J.F.

    1988-01-01

    The March 1979 accident at Three Mile Island Unit 2 (TMI-2) damaged the core of the reactor. One of the major cleanup activities involves removal of the damaged core from the reactor and transporting it from the TMI-2 site near Middletown, Pennsylvania, to the Idaho National Engineering Laboratory (INEL) near Idaho Falls, Idaho. Removal and transport of the damaged core necessitated the development of much specialized equipment. This paper focuses on the role quality assurance (QA) played in the design, fabrication, acceptance, and use of three important pieces of core debris removal and transportation equipment: (1) the core boring machine, (2) the fuel debris canisters, (3) the NuPac 125-B rail cask and handling equipment

  18. Assessment of TMI-2 I and Cs chemistry during core degradation

    International Nuclear Information System (INIS)

    Cronenberg, A.W.; Langer, S.

    1987-09-01

    Data from the TMI-2 accident has shown that only small amounts of iodine and cesium escaped the plant, on the order of tens of curies. A detailed investigation of fission product I and Cs behavior during the core degradation phase of the TMI-2 accident was initiated. Specifically, analyses are presented of I and Cs release from the severely damaged fuel, chemical reaction processes with the steam/hydrogen effluent, and transport/deposition behavior within the primary coolant system. Results indicate that, although elemental I and Cs release from fuel can be expected, the effluent chemical environment was such as to favor predominantly CsI and CsOH formation in the high-temperature effluent. Subsequent CsI reaction with boric acid released with reflood water is predicted to have resulted in partial conversion of CsI to Cs-borate and HI. Residual transport of CsI is predicted to condense and chemisorb in the upper reactor plenum, forming additional HI in the process. HI is, therefore, assessed to be the principal form of iodine transport leaving the reactor vessel during core degradation, which was subsequently dissolved in water to produce iodide ions in solution with coolant. 19 figs., 24 tabs

  19. Reactor building for a nuclear reactor

    International Nuclear Information System (INIS)

    Haidlen, F.

    1976-01-01

    The invention concerns the improvement of the design of a liner, supported by a latticed steel girder structure and destined for guaranteeing a gastight closure for the plant compartments in the reactor building of a pressurized water reactor. It is intended to provide the steel girder structure on their top side with grates, being suited for walking upon, and to hang on their lower side diaphragms in modular construction as a liner. At the edges they may be sealed with bellows in order to avoid thermal stresses. The steel girder structure may at the same time serve as supports for parts of the steam pipe. (RW) [de

  20. Criticality safety for TMI-2 canister storage at INEL

    International Nuclear Information System (INIS)

    Jones, R.R.; Briggs, J.B.; Ayers, A.L. Jr.

    1986-01-01

    Canisters containing Three Mile Island Unit 2 (TMI-2) core debris will be researched, stored, and prepared for final disposition at the Idaho National Engineering Laboratory (INEL). The canisters will be placed into storage modules and assembled into a storage rack, which will be located in the Test Area North (TAN) storage pool. Criticality safety calculations were made (a) to ensure that the storage rack is safe for both normal and accident conditions and (b) to determine the effects of degradation of construction materials (Boraflex and polyethylene) on criticality safety

  1. Evaluation of the Three Mile Island Unit 2 reactor building decontamination process

    International Nuclear Information System (INIS)

    Dougherty, D.; Adams, J.W.

    1983-08-01

    Decontamination activities from the cleanup of the Three Mile Island Unit 2 Reactor Building are generating a variety of waste streams. Solid wastes being disposed of in commercial shallow land burial include trash and rubbish, ion-exchange resins (Epicor-II) and strippable coatings. The radwaste streams arising from cleanup activities currently under way are characterized and classified under the waste classification scheme of 10 CFR Part 61. It appears that much of the Epicor-II ion-exchange resin being disposed of in commerical land burial will be Class B and require stabilization if current radionuclide loading practices continue to be followed. Some of the trash and rubbish from the cleanup of the reactor building so far would be Class B. Strippable coatings being used at TMI-2 were tested for leachability of radionuclides and chelating agents, thermal stability, radiation stability, stability under immersion and biodegradability. Actual coating samples from reactor building decontamination testing were evaluated for radionuclide leaching and biodegradation

  2. Evaluation of the Three Mile Island Unit 2 reactor building decontamination process

    Energy Technology Data Exchange (ETDEWEB)

    Dougherty, D.; Adams, J. W.

    1983-08-01

    Decontamination activities from the cleanup of the Three Mile Island Unit 2 Reactor Building are generating a variety of waste streams. Solid wastes being disposed of in commercial shallow land burial include trash and rubbish, ion-exchange resins (Epicor-II) and strippable coatings. The radwaste streams arising from cleanup activities currently under way are characterized and classified under the waste classification scheme of 10 CFR Part 61. It appears that much of the Epicor-II ion-exchange resin being disposed of in commerical land burial will be Class B and require stabilization if current radionuclide loading practices continue to be followed. Some of the trash and rubbish from the cleanup of the reactor building so far would be Class B. Strippable coatings being used at TMI-2 were tested for leachability of radionuclides and chelating agents, thermal stability, radiation stability, stability under immersion and biodegradability. Actual coating samples from reactor building decontamination testing were evaluated for radionuclide leaching and biodegradation.

  3. Method of constructing reactor buildings

    International Nuclear Information System (INIS)

    Hyuga, Takenori; Nagai, Fumio; Akutsu, Masayoshi.

    1985-01-01

    Purpose: To shorten the construction period for LMFBR type reactors, as well as smoothly introduce high pressure steams generated in concretes upon loss of coolant accidents to the outside of the system. Method: After disposing a liner plate as a chamber lining of reactor buildings, heat insulation materials having steam discharge channels at the outer surface are attached to the outside of the liner plate and, further, an organic films are disposed to the outside of the heat insulation materials. Then, concretes are spiked to the outside of the organic films using the liner plate and the heat insulation material as the mold for concretes. In this way, the construction period can be shortened by utilizing the liner plate and the heat insulation materials as the mold for concretes, as well as steams at high temperature resulted in the concretes upon loss of coolant accidents can smoothly be discharged to the outside of the system. (Moriyama, K.)

  4. Environmental assessment for decontamination of the Three Mile Island Unit 2 reactor building atmosphere. Draft NRC staff report for public comment

    International Nuclear Information System (INIS)

    1980-03-01

    The krypton-85 (Kr-85) released to the reactor building during the accident at TMI-2 must be removed from the reactor building in order to permit greater access to the building than is currently possible. The gases currently in the building emit sufficient radiation (1.2 rem/hr total body, 150 rad/hr skin dose) that occupation of the reactor building is severely limited even with protective clothing. Greater access is likely to be necessary to maintain instrumentation and equipment required to keep the reactor in a safe shutdown condition. In addition greater access would facilitate the gathering of data needed for planning the building decontamination program. An additional consideration is that prolonged enclosure of the Kr-85 within the building greatly increases the risk of its successive uncontrolled releases to the outside environment. The staff's evaluation of alternative methods for removing the krypton shows that each could be implemented with little risk to the health and safety of the public. The reactor building purge system, charcoal adsorption system, gas compression, selective absorption process system, and cryogenic processing system could each be operated to keep levels of airborne radioactive materials to unrestricted areas in compliance with the requirements of 10 CFR Part 20, and the design objectives of Appendix 1 to 10 CFR Part 50 of the Commission's regulations, and with the applicable requirements of 40 CFR Part 190.10

  5. Assessment of Extent and Degree of Thermal Damage to Polymeric Materials in the Three Mile Island Unit 2 Reactor Building

    International Nuclear Information System (INIS)

    Alvares, N. J.

    1984-02-01

    Thermal damage to susceptible materials in accessible regions of the TMI-2 reactor building shows damage-distribution patterns that indicate non-uniform intensity of exposure. No clear explanation for non-uniformity is found in existing evidence; e.g., in some regions a lack of thermally susceptible materials frustrates analysis. Elsewhere, burned materials are present next to materials that seem similar but appear unscathed-leading to conjecture that the latter materials preferentially absorb water vapor during periods of high local steam concentration. Most of the polar crane pendant shows heavy burns on one half of its circumferential surface. This evidence suggests that the polar crane pendant side that experienced heaviest burn damage was exposed to intense radiant energy from a transient fire plume in the reactor containment volume. Tests and simple heat-transfer calculations based on pressure and temperature records from the accident show that the atmosphere inside the reactor building was probably 8% hydrogen in air, a value not inconsistent with the extent of burn damage. Burn-pattern geography indicates uniform thermal exposure in the dome volume to the 406-ft level (about 6 ft below the polar crane girder), partial thermal exposure in the volume between the 406- and 347-ft levels as indicated by the polar crane cable, and lack of damage to most thermally susceptible materials in the west quadrant of the reactor building; some evidence of thermal exposure Is seen in the free volume between the 305- and 347-ft levels. (author)

  6. Calculations to estimate the margin to failure in the TMI-2 vessel

    International Nuclear Information System (INIS)

    Stickler, L.A.; Rempe, J.L.; Chavez, S.A.; Thinnes, G.L.; Snow, S.D.; Witt, R.J.; Corradini, M.L.; Kos, J.A.

    1994-03-01

    As part of the OECD-sponsored Three Mile Island Unit 2 (TMI-2) Vessel Investigation Project (VIP), margin-to-failure calculations for mechanisms having the potential to threaten the integrity of the vessel were performed to improve understanding of events that occurred during the TMI-2 accident. Analyses considered four failure mechanisms: tube rupture, tube ejection, global vessel failure, and localized vessel failure. Calculational input was based on data from the TMI-2 VIP examinations of the vessel steel samples, the instrument tube nozzles, and samples of the hard layer of debris found on the TMI-2 vessel lower head. Sensitivity studies were performed to investigate the uncertainties in key parameters for these analyses

  7. A Brief Review of Past INL Work Assessing Radionuclide Content in TMI-2 Melted Fuel Debris: The Use of 144Ce as a Surrogate for Pu Accountancy

    Energy Technology Data Exchange (ETDEWEB)

    D. L. Chichester; S. J. Thompson

    2013-09-01

    This report serves as a literature review of prior work performed at Idaho National Laboratory, and its predecessor organizations Idaho National Engineering Laboratory (INEL) and Idaho National Engineering and Environmental Laboratory (INEEL), studying radionuclide partitioning within the melted fuel debris of the reactor of the Three Mile Island 2 (TMI-2) nuclear power plant. The purpose of this review is to document prior published work that provides supporting evidence of the utility of using 144Ce as a surrogate for plutonium within melted fuel debris. When the TMI-2 accident occurred no quantitative nondestructive analysis (NDA) techniques existed that could assay plutonium in the unconventional wastes from the reactor. However, unpublished work performed at INL by D. W. Akers in the late 1980s through the 1990s demonstrated that passive gamma-ray spectrometry of 144Ce could potentially be used to develop a semi-quantitative correlation for estimating plutonium content in these materials. The fate and transport of radioisotopes in fuel from different regions of the core, including uranium, fission products, and actinides, appear to be well characterized based on the maximum temperature reached by fuel in different parts of the core and the melting point, boiling point, and volatility of those radioisotopes. Also, the chemical interactions between fuel, fuel cladding, control elements, and core structural components appears to have played a large role in determining when and how fuel relocation occurred in the core; perhaps the most important of these reaction appears to be related to the formation of mixed-material alloys, eutectics, in the fuel cladding. Because of its high melting point, low volatility, and similar chemical behavior to plutonium, the element cerium appears to have behaved similarly to plutonium during the evolution of the TMI-2 accident. Anecdotal evidence extrapolated from open-source literature strengthens this logical feasibility for

  8. Fuel models and results from the TRAC-PF1/MIMAS TMI-2 accident calculation

    International Nuclear Information System (INIS)

    Schwegler, E.C.; Maudlin, P.J.

    1983-01-01

    A brief description of several fuel models used in the TRAC-PF1/MIMAS analysis of the TMI-2 accident is presented, and some of the significant fuel-rod behavior results from this analysis are given. Peak fuel-rod temperatures, oxidation heat production, and embrittlement and failure behavior calculated for the TMI-2 accident are discussed. Other aspects of fuel behavior, such as cladding ballooning and fuel-cladding eutectic formation, were found not to significantly affect the accident progression

  9. Probabilistic Dose Assessment from SB-LOCA Accident in Ujung Lemahabang Using TMI-2 Source Term

    Directory of Open Access Journals (Sweden)

    Sunarko

    2017-01-01

    Full Text Available Probabilistic dose assessment and mapping for nuclear accident condition are performed for Ujung Lemahabang site in Muria Peninsula region in Indonesia. Source term is obtained from Three-Mile Island unit 2 (TMI-2 PWR-type SB-LOCA reactor accident inverse modeling. Effluent consisted of Xe-133, Kr-88, I-131, and Cs-137 released from a 50 m stack. Lagrangian Particle Dispersion Method (LPDM and 3-dimensional mass-consistent wind field are employed to obtain surface-level time-integrated air concentration and spatial distribution of ground-level total dose in dry condition. Site-specific meteorological data is obtained from hourly records obtained during the Site Feasibility Study period in Ujung Lemahabang. Effluent is released from a height of 50 meters in uniform rate during a 6-hour period and the dose is integrated during this period in a neutrally stable atmospheric condition. Maximum dose noted is below regulatory limit of 1 mSv and radioactive plume is spread mostly to the W-SW inland and to N-NE from the proposed plant to Java Sea. This paper has demonstrated for the first time a probabilistic analysis method for assessing possible spatial dose distribution, a hypothetical release, and a set of meteorological data for Ujung Lemahabang region.

  10. Hanford B Reactor Building Hazard Assessment Report

    International Nuclear Information System (INIS)

    Griffin, P. W.

    1999-01-01

    The 105-B Reactor (hereinafter referred to as B Reactor) is located in the 100 Area of the Hanford Site near Richland, Washington. The B Reactor is one of nine plutonium production reactors that were constructed in the 1940s during the Cold War Era. Construction of the B Reactor began June 7, 1943, and operation began on September 26, 1944. The Environmental Restoration Contractor was requested by RL to provide an assessment/characterization of the B Reactor building to determine and document the hazards that are present and could pose a threat to the environment and/or to individuals touring the building. This report documents the potential hazards, determines the feasibility of mitigating the hazards, and makes recommendations regarding areas where public tour access should not be permitted

  11. Radiation and health effects. A report on the TMI-2 accident and related health studies

    International Nuclear Information System (INIS)

    1986-08-01

    On March 28, 1979, the Unit 2 reactor at the Three Mile Island (TMI) Nuclear Station was severely damaged by an accident. Radioactivity was discharged to the environment resulting in a small amount of radiation exposure to the public. Continuing concerns by some members of the communities around TMI about the potential radiation-induced health effects prompted GPU Nuclear Corporation to examine the information gathered from the accident investigation in the context of our current knowledge of radiation and its effects on human health. Although this report deals with technical matters, the information is presented in a manner that can be understood by those who do not have scientific backgrounds. This report is divided into three major sections. The first section provides an overview of the past 80 years of relevant research on the subject of radiation and its effects on human health. During that time, scientists and physicians throughout the world have studied hundreds of thousands of individuals exposed to radiation from medical and occupational sources and from nuclear weapons explosions. Epidemiologic studies of humans, such as the Japanese survivors of the atomic bomb, have established that following exposure to large doses of radiation, certain health effects, including cancer, can be observed. Radiation-induced health effects from low doses of radiation, such as those associated with the TMI-2 accident, appear infrequently, if at all, and are identical and, therefore, indistinguishable from similar health effects which occur normally. For example, cancers induced by radiation are indistinguishable from those occurring spontaneously or normally. It is not possible, therefore, for scientists to determine directly whether radiation-induced health effects at low doses occur at all; such observations can only be inferred by statistical methods. The second section of this report provides a brief description of the TMI-2 accident. Most of the radioactivity from the

  12. ITER [International Thermonuclear Experimental Reactor] reactor building design study

    International Nuclear Information System (INIS)

    Thomson, S.L.; Blevins, J.D.; Delisle, M.W.

    1989-01-01

    The International Thermonuclear Experimental Reactor (ITER) is at the midpoint of a two-year conceptual design. The ITER reactor building is a reinforced concrete structure that houses the tokamak and associated equipment and systems and forms a barrier between the tokamak and the external environment. It provides radiation shielding and controls the release of radioactive materials to the environment during both routine operations and accidents. The building protects the tokamak from external events, such as earthquakes or aircraft strikes. The reactor building requirements have been developed from the component designs and the preliminary safety analysis. The equipment requirements, tritium confinement, and biological shielding have been studied. The building design in progress requires continuous iteraction with the component and system designs and with the safety analysis. 8 figs

  13. Vibration-damping structure for reactor building

    International Nuclear Information System (INIS)

    Kuno, Toshio; Iba, Chikara; Tanaka, Hideki; Kageyama, Mitsuru

    1998-01-01

    In a damping structure of a reactor building, an inner concrete body and a reactor container are connected by way of a vibration absorbing member. As the vibration absorbing member, springs or dampers are used. The inner concrete body and the reactor container each having weight and inherent frequency different from each other are opposed displaceably by way of the vibration absorbing member thereby enabling to reduce seismic input and reduce shearing force at least at leg portions. Accordingly, seismic loads are reduced to increase the grounding rate of the base thereby enabling to satisfy an allowable value. Therefore, it is not necessary to strengthen the inner concrete body and the reactor container excessively, the amount of reinforcing rods can be reduced, and the amount of a portion of the base buried to the ground can be reduced thereby enabling to constitute the reactor building easily. (N.H.)

  14. Modelling of the liquid level during uncovering of the TMI-2 core

    International Nuclear Information System (INIS)

    Ardron, K.H.; Cain, D.G.

    1980-01-01

    Calculation of the path followed by the falling liquid level in the core is an important part of the analysis of fuel heat-up, and the damage sustained in the accident at TMI-2. This paper describes the development and application of a model for predicting liquid level in the TMI-2 core during uncovering. The calculational method is shown to give a satisfactory representation of level height data obtained in TMI boil-down simulation tests in the Semi-scale facility. The model is combined with a best estimate of make-up flow to the TMI-2 core to calculate the level trajectory during core uncovering. Results are found to be sensitive to the assumed degree of steam condensation on the make-up flow

  15. Quality Assurance in the removal and transport of the TMI-2 core

    International Nuclear Information System (INIS)

    Hayes, G.R.; Marsden, J.F.

    1988-01-01

    EG ampersand G Idaho, acting on behalf of the US Department of Energy (DOE), is cooperating with the owner of the TMI-2 plant, General Public Utilities Nuclear (GPUN), in the removal and transport of the damaged TMI-2 core to the Idaho National Engineering Laboratory (INEL) near Idaho Falls, Idaho. Quality Assurance (QA) played an important role in the removal and transport of the damaged TMI-2 core. To illustrate, the authors have chosen to discuss some of the important quality assurance techniques utilized in the design, fabrication, acceptance, and use of the three different types of equipment; the core boring machine, the core debris canisters, and the transport casks. Rather than a thorough discussion of the QA aspects of each task, the authors have purposely chosen to present only the key applications of quality assurance principles and methodology unique to each piece of equipment. The intent of this approach is to effectively communicate the importance of ''task teamwork'' in QA

  16. Implementation of remote equipment at TMI-2 [Three Mile Island Unit 2

    International Nuclear Information System (INIS)

    Giefer, D.; Jeffries, A.B.

    1988-01-01

    Each of the remote vehicles in use, or planned for use, at Three Mile Island Unit 2 (TMI-2) during the period from 1984 to the present had certain distinct common features. These were proven to be desirable for remote application in the TMI-2 environment. Proper implementation requires consideration of the following: control systems, rigging systems, power supplies, operator/support interface, maintenance concerns, viewing systems, contamination control, and communications. Design and component fabrication of these features allowed deployment of each of the remote devices. This paper discusses these systems and their impact for the use of remote mobile equipment at TMI-2. In most cases, the means of implementation dictated the design features of the devices

  17. Pressure suppression device for nuclear reactor building

    International Nuclear Information System (INIS)

    Ikegame, Noboru.

    1992-01-01

    In a nuclear reactor building, there are disposed cooling coils connected to an air supply duct at the outside of the building, an air supply blower, an air supply duct having the top end opened, an exhaustion duct having the top end opened and a bypassing pipeline interposed between the exhaustion duct and the air supply duct on the side of the inlet of the cooling coils. In the reactor building, when a radioactive material leakage accident should occur, an isolation valve is closed to isolate the building from the outside. Further, bypassing isolation valve is opened to form a closed cooling circuit by the cooling coils, the air supply blower and the air supply duct, the exhaustion duct and the bypassing pipeline in the reactor building. With such a constitution, since air as the atmosphere in the building is circulated through the closed cooling circuit and cooled by the cooling coils, the temperature is not elevated. Accordingly, since the pressure elevation of the atmosphere in the building is suppressed, the atmosphere containing radioactive materials do not flow out of the building. (I.N.)

  18. Interim report on the TMI-2 purification filter examination

    International Nuclear Information System (INIS)

    Mason, R.E.; Hobbins, R.R.; Cook, B.A.; MacDonald, P.E.

    1983-02-01

    Filters from the purification/makeup system of the Three Mile Island Unit 2 Reactor were examined after the March 28, 1979, accident to determine the character of the debris transported to the filters. The general condition of the filters is presented. Material was removed from the filters and examined. The elemental and radionuclide makeup of the debris is discussed. Distribution of particle size and shape is presentd for some of the material examined. This is an interim report. When the investigation is completed, another report summarizing all of the data will be issued

  19. Tellurium release and deposition during the TMI-2 accident

    International Nuclear Information System (INIS)

    Vinjamuri, K.; Osetek, D.J.; Hobbins, R.R.; Jessup, J.S.

    1984-09-01

    The estimated behavior of tellurium during and after the accident at the Three Mile Island Unit-2 is presented. The behavior is based on all available measurement data for /sup 129m/Te, 132 Te, stable tellurium ( 126 Te, 128 Te and 130 Te), and best estimate calculations of tellurium release and transport. The predicted release was calculated using current techniques that relate release rate to fuel temperature and holdup of tellurium in zircaloy until significant oxidation occurs. The calculated release fraction was low, approx. 7%, but the total measured release for samples analyzed to date is about 5.8%. Of the measured tellurium about 2.4, 1.8, 0.88, 0.42, 0.17 and 0.086% of core inventory were in the containment sump water, upper plenum assembly surfaces, containment solids in the sump water, makeup and purification demineralizer, containment inside surface, and the reactor primary coolant, respectively. A significant fraction (54%) of the tellurium calculated to be retained on the upper plenum surfaces (4.61% of the core inventory) was deposited during the high pressure injection of coolant at about 200 min after the reactor scram. Comparison of tellurium behavior with in-pile and out-of-pile tests strongly suggests that zircaloy holds tellurium until significant cladding oxidation occurs

  20. Reactor building assembly and method of operation

    International Nuclear Information System (INIS)

    Fennern, L.E.; Caraway, H.A.; Hsu, Li C.

    1993-01-01

    A reactor building assembly is described comprising: a reactor pressure vessel containing a reactor core for generating heat in the form of steam; a containment vessel enclosing said pressure vessel; a first enclosure surrounding said containment vessel and spaced laterally therefrom to define a first chamber there between, and having a top and a bottom; a second enclosure surrounding said first enclosure and spaced laterally therefrom to define a second chamber there between, and having a top and a bottom; a building inlet for receiving into said second chamber fresh air from outside said second enclosure; a building outlet for discharging stale air from said first chamber; a transfer duct disposed through said first enclosure selectively joining in flow communication said first and second chambers; said building inlet being disposed at said second enclosure top, said building outlet being disposed at said first enclosure top, and said transfer duct being disposed adjacent said first enclosure bottom for allowing said fresh air to flow downwardly by gravity through said second chamber and through said transfer duct into said first chamber for cooling said first chamber, said stale air flowing upwardly by natural buoyancy for discharger from said first chamber through said building outlet; an exhaust stack disposed above said building outlet and in flow communication therewith for channeling upwardly said stale air from said first chamber for discharge into the surrounding environs; and a passive first driving means for increasing flow of said stale air from said building outlet comprising: an isolation pool containing isolation water; an isolation condenser disposed in said isolation pool, and joined in flow communication with said reactor pressure vessel for receiving primary steam therefrom, said primary steam being cooled in said isolation condenser for heating said isolation water to generate secondary steam

  1. Analysis of the Three Mile Island (TMI-2) hydrogen burn

    International Nuclear Information System (INIS)

    Henrie, J.O.; Postma, A.K.

    1982-07-01

    A study of recorded temperatures and pressures was made. Hydrogen concentrations were calculated. Postburn average ambient temperatures versus time were calculated. Average temperatures were calculated for the region above elevation 347, below elevation 347, and within the D-shield compartments. Results were: Prior to the burn, the hydrogen was well mixed with the containment air. Average hydrogen concentration was 7.9%, wet basis. The hydrogen burn occurred at all three levels in the containment. The burn was initiated somewhere in the lowest level, probably on the west side. The burn time was about 12 s. About 3570 m 3 or 126,000 ft 3 , 160 kg moles or 319 kg of hydrogen burned. Containment gas temperatures in the flame front were 760 0 C (1400 0 F). Average containment gas temperature at end was 660 0 C (1220 0 F). The gas temperatures decreased much faster below elevation 347 than above elevation 347. The average temperature rise was only about 1.2 0 C (2.2 0 F) as a result of the hydrogen burn. Considerably more energy came from the hot water and steam vented. The burn damage observed was predominantly at the upper elevations and on the north, east, and south quadrants. 1.1% hydrogen remained in the containment after the burn. Venting of the reactor cooling system during the hour following the burn added an additional 0.6%. Hydrogen concentrations increased from this 1.7% to about 2.2% between March 30 and April 2. One of two Rockwell hydrogen recombiners removed 112 kg of hydrogen. A total of 459 kg of hydrogen gas were accounted for. Assuming that 432 kg were generated by the zirconium-steam reaction, 9850 kg Zr would have been oxidized

  2. Computer aided analysis of reactor containment building

    International Nuclear Information System (INIS)

    Abhat, O.B.

    1985-01-01

    The paper presents a computerized, structural analysis of the Reactor Containment Building (RCB) of a liquid metal nuclear power plant. Emphasis is placed on developing techniques for large 3-Dimensional, structural models using interactive computer graphics, analyzing the structure by NASTRAN and a simplified analytical approach to account for non-linear, temperature-dependent, material properties at elevated temperatures caused by a sodium spill from a Design Basis Accident (DBA). (orig.)

  3. Criteria development of remotely controlled mobile devices for TMI-2 [Three Mile Island Unit 2

    International Nuclear Information System (INIS)

    Fillnow, R.; Bengel, P.; Giefer, D.

    1988-01-01

    Since 1982, GPU Nuclear Corporation has used a series of remote mobile devices for data collection and cleanup of highly contaminated areas in the Three Mile Island Unit 2 (TMI-2) nuclear facilities. This paper describes these devices and the general criteria established for their design. Until 1984, the remote equipment used at TMI was obtained from industry sources. This included devices called SISI, FRED, and later LOUIE-1. Following 1984, the direction was to obtain custom-made devices to assure a design that would be more appropriate for the TMI-2 environment. Along with this approach came more detailed criteria and a need for a thorough understanding of the task to be accomplished by the devices. The following families of equipment resulted: (1) remote reconnaissance vehicles (RRVs), (2) the LOUIE family, and (3) remote working vehicle (RWV) family

  4. TMI-2 [Three Mile Island Unit 2] primary coolant mass flowrate data report

    International Nuclear Information System (INIS)

    McCormick, R.D.

    1986-12-01

    This is a report on the preparation of data from the TMI-2 primary coolant mass flowrate meters for inclusion into the TMI Data Base. The sources of the as-recorded data are discussed, and a description of the instrument is given. An explanation is given of how corrections were made to the as-recorded data and how the uncertainties were calculated. The identifiers attached to each data set in the TMI Data Base are given

  5. TMI-2 source and intermediate range neutron flux monitors data report

    International Nuclear Information System (INIS)

    McCormick, R.D.

    1986-03-01

    This is a report on the preparation of data from the TMI-2 excore source and intermediate range neutron flux monitors for inclusion into the TMI Data Base. The sources of the as-recorded data are discussed as well as the process of transforming these data into digital form. The corrections to the as-recorded data are given and the data quality classification and uncertainty are established. The identifiers attached to each data set in the TMI Data Base are given

  6. Seismic calculations for underground reactor buildings

    International Nuclear Information System (INIS)

    Altes, J.; Koschmieder, D.

    1977-08-01

    Embedding the buildings in soil changes their seismic response behaviour as compared to surface buildings, i.e. higher stiffness and increased radiation damping is attained. Finite element models are best suited for determinig the effects of embedment and of layered subsoil. The code used was the LUSH2-programme, which is applicable to 2-dimensional problems and provides an approximate treatment for non-linear dynamic soil behaviour. For embedded buildings there is a good agreement between 2- and 3-dimensional models of the response for points below the soil surface. It is therefore permissible to use the less costly 2-dimensional programmes. To simulate earthquake, three different acceleration-time histories, derived from actual measurements and from artificial synthesis, with differing response spectra were fed in. The soil characteristics assumed are applicable to a representative site in Germany. Three different types of models were examined, using analytical models with only a few elements for parametric studies and with up to 716 elements for more precise calculations. A comparison was made between the semi-embedment, the total embedment, and installation of the reactor building above-ground. (orig.) [de

  7. Revision to ANSI/ANS 3.1 1978: resulting from TMI-2

    International Nuclear Information System (INIS)

    Palmer, F.A.

    1981-01-01

    The personnel errors which occurred at TMI-2 brought forth several areas of weaknesses in personnel selection, qualification and training that impacted ANS-3. As a result, the ANS-3 Committee started working on a revision to ANSI/ANS 3.1-1978 Standard in May 1979. In July 1979 the first set of official recommendations was issued in NUREG 0578. Due to the interim nature of these regulations some interpretation of the intent of these recommendations had to be made and a basis developed to justify changes to the standard

  8. Reactor and turbine building layout of the high performance light water reactor

    International Nuclear Information System (INIS)

    Bittermann, D.

    2010-01-01

    Based on the information generated within the European funded project ''High Per-formance Light Water Reactor Phase 2'', a general plant layout has been developed. The central building is the reactor building, in which the containment and safety sys-tems are located. The reactor building is with app. 90.000 m 3 considerably smaller compared to other BWR buildings, thus providing a huge potential for cost savings. The turbine building with app 250,000 m 3 is of approximately the same size like for existing BWRs. (orig.)

  9. Site dose calculations for the INEEL/TMI-2 storage facility

    International Nuclear Information System (INIS)

    Jones, K.B.

    1997-01-01

    The U.S. Department of Energy (DOE) is licensing an independent spent-fuel storage installation (ISFSI) for the Three Mile Island unit 2 (TMI-2) core debris to be constructed at the Idaho Chemical Processing Plant (ICPP) site at the Idaho National Engineering and Environmental Laboratory (INEEL) using the NUHOMS spent-fuel storage system. This paper describes the site dose calculations, performed in support of the license application, that estimate exposures both on the site and for members of the public. These calculations are unusual for dry-storage facilities in that they must account for effluents from the system in addition to skyshine from the ISFSI. The purpose of the analysis was to demonstrate compliance with the 10 CFR 20 and 10 CFR 72.104 exposure limits

  10. Tellurium chemistry, tellurium release and deposition during the TMI-2 accident

    International Nuclear Information System (INIS)

    Vinjamuri, K.; Sallach, R.A.; Osetek, D.J.; Hobbins, R.R.; Akers, D.W.

    1985-08-01

    This report presents the chemistry and estimated behavior of tellurium during and after the accident at Three Mile Island Unit-2. The discussion of tellurium behavior is based on all available measurement data for /sup 129 m/Te, 132 Te, stable tellurium ( 126 Te, 128 Te, and 130 Te), and best estimate calculations of tellurium release and transport. Results from Oak Ridge National Laboratory (ORNL) tests, Power Burst Facility (PBF) Severe Fuel Damage Tests at Idaho National Engineering Laboratory (INEL) and SASCHA tests from Karlsruhe, W. Germany are compared with calculated release fractions and samples taken from TMI Unit-2. It is concluded that very little tellurium was released and transported from the TMI-2 core, probably as a result of holdup by zircaloy cladding and other structural materials. 39 refs., 24 figs., 17 tabs

  11. Final report on the in situ testing of electrical components and devices at TMI-2

    International Nuclear Information System (INIS)

    Soberano, F.T.

    1984-06-01

    A total of 88 electrical components and devices were in situ tested. Of these, 11 totally failed and 21 suffered degradation that varied from mild to severe. The equipment that failed or incurred severe degradation was located in areas of known environmental extremes. Several motor operated valves in the Reactor Building basement failed because of submersion in water. Others severely degraded from contamination tracking, resulting in the alteration of their circuit electrical characteristics - a circumstance that could compromise their designed function. One backup oil lift pump motor for a reactor coolant pump motor, although located well above the Reactor Building basement high water mark, failed because of a break in its armature and field circuits; this failure was surmised to be a result of corrosion. The limit switch of a Class 1E solenoid valve likewise failed due to moisture intrusion. Components that noticeably degraded exhibited anomalies, likely due to the incursion of moisture, that varied from high capacitance to increased circuit resistance. The effect of the other degenerating conditions that existed during the accident, such as high temperature, high radiation levels, and the hydrogen burn, could not be evaluated individually or synergistically

  12. Development of the APR+ Reactor Containment Building General Arrangement (GA)

    International Nuclear Information System (INIS)

    Moon, Hyung Keun; Kim, Dae Hun; Park, Young Sheop

    2011-01-01

    The general arrangement (GA) drawing of a nuclear power plant is the most basic drawing, containing all of the plant equipment, systems, and rooms. Therefore, it should be issued at an early design stage to provide the contours of the overall plant structure. This type of drawing is typically used widely throughout the design stages. The development project of the APR+ (Advanced Power Reactor+), as a model to succeed the APR1400 (Advanced Power Reactor 1400) design, has its own GA that encompasses all of its power buildings. This drawing was developed starting in October of 2009. Among several of the buildings in this design, the Reactor Containment Building (RCB), Auxiliary Building (AB), and EDG (Emergency Diesel Generator) Building are some the most important buildings to produce electricity, and to protect against undesirable radiation emissions. This paper focuses on the design characteristics of the general arrangement of the RCB

  13. Ventilation system in the RA reactor building - design specifications

    International Nuclear Information System (INIS)

    Badrljica, R.

    1984-09-01

    Protective role of the ventilation system of nuclear facilities involve construction of ventilation barriers which prevent release of radioactive particulates or gases, elimination od radioactive particulates and gases from the air which is released from contaminated zones into the reactor environment. Ventilation barriers are created by dividing the building into a number of ventilation zones with different sub pressure compared to the atmospheric pressure. The RA reactor building is divided into four ventilation zones. First zone is the zone of highest risk. It includes reactor core with horizontal experimental channels, underground rooms of the primary coolant system (D 2 O), helium system, hot cells and the space above the the reactor core. Second zone is the reactor hall and the room for irradiated fuel storage. The third zone includes corridors in the basement, ground floor and first floor where the probability of contamination is small. The fourth zone includes the annex where the contamination risk is low. There is no have natural air circulation in the reactor building. Ventilators for air input and outlet maintain the sub pressure in the building (pressure lower than the atmospheric pressure). This prevents release of radioactivity into the atmosphere [sr

  14. Thermal interaction of core melt debris with the TMI-2 baffle, core-former, and lower head structures

    International Nuclear Information System (INIS)

    Cronenberg, A.W.; Tolman, E.L.

    1987-09-01

    Recent inspection of the TMI-2 core-former baffle walls (vertical), former plates (horizontal), and lower plenum has been conducted to assess potential damage to these structures. Video observations show evidence of localized melt failure of the baffle walls, whereas fiberoptics data indicate the presence of resolidified debris on the former plates. Lower plenum inspection also confirms the presence of 20 tons or more of core debris in the lower plenum. These data indicate massive core melt relocation and the potential for melt attack on vessel structural components. This report presents analyses aimed at developing an understanding of melt relocation behavior and damage progression to TMI-2 vessel components. Thermal analysis indicates melt-through of the baffle plates, but maintenance of structural integrity of the former plates and lower head. Differences in the damage of these structures is attributed largely to differences in contact time with melt debris and pressure of water. 29 refs., 17 figs., 9 tabs

  15. Nonlinear analysis of a reactor building for airplane impact loadings

    International Nuclear Information System (INIS)

    Zimmermann, T.; Rodriguez, C.; Rebora, B.

    1981-01-01

    The purpose is to analyze the influence of material nonlinear behavior on the response of a reinforced concrete reactor building and on equipment response for airplane impact loadings. Two analyses are performed: first, the impact of a slow-flying commercial airplane (Boeing 707), then the impact of a fast flying military airplane (Phantom). (orig./HP)

  16. Overview of the Westinghouse Small Modular Reactor building layout

    Energy Technology Data Exchange (ETDEWEB)

    Cronje, J. M. [Westinghouse Electric Company LLC, Centurion (South Africa); Van Wyk, J. J.; Memmott, M. J. [Westinghouse Electric Company LLC, Cranberry Township, PA (United States)

    2012-07-01

    The Westinghouse Small Modular Reactor (SMR) is an 800 MWt (>225 MWe) integral pressurized water reactor (iPWR), in which all of the components typically associated with the nuclear steam supply system (NSSS) of a nuclear power plant are incorporated within a single reactor pressure vessel. This paper is the third in a series of four papers, which describe the design and functionality of the Westinghouse SMR. It focuses in particular upon the plant building layout and modular design of the Westinghouse SMR. In the development of small modular reactors, the building layout is an area where the safety of the plant can be improved by applying new design approaches. This paper will present an overview of the Westinghouse SMR building layout and indicate how the design features improve the safety and robustness of the plant. The Westinghouse SMR is designed with no shared systems between individual reactor units. The main buildings inside the security fence are the nuclear island, the rad-waste building, the annex building, and the turbine building. All safety related equipment is located in the nuclear island, which is a seismic class 1 building. To further enhance the safety and robustness of the design, the reactor, containment, and most of the safety related equipment are located below grade on the nuclear island. This reduces the possibility of severe damage from external threats or natural disasters. Two safety related ultimate heat sink (UHS) water tanks that are used for decay heat removal are located above grade, but are redundant and physically separated as far as possible for improved safety. The reactor and containment vessel are located below grade in the center of the nuclear island. The rad-waste and other radioactive systems are located on the bottom floors to limit the radiation exposure to personnel. The Westinghouse SMR safety trains are completely separated into four unconnected quadrants of the building, with access between quadrants only allowed

  17. Life management for a non replaceable structure: the reactor building

    International Nuclear Information System (INIS)

    Torres, V.; Francia, L.

    1998-01-01

    Phase 1 of UNESA N.P.P. Lifetime Management Project identified and ranked important components, relative to plant life management. The list showed the Reactor Containment Structure in the third position, and thirteen concrete structures were among the list top twenty. Since the Reactor Containment Building, together with the Reactor Vessel, is the only non-replaceable plant component, and has a big impact on the plant technical life, there is an increasing interest on understanding its behavior to maintain structural integrity. This paper presents: a) IAEA (International Atomic Energy Agency) Coordinated Research Program experiences and studies. Under this Program, international experts address the most frequent degradation mechanisms affecting the containment building. b) IAEA Aging Management Program adapted to our plants. The paper addresses the aging mechanisms affecting the concrete structures, reinforcing steel and prestress systems as well as the aging management programs and the mitigation and control methods. Finally, this paper presents a new module called STRUCTURES, included in phase 2 of the above mentioned project, which will monitor and document the different aging mechanisms and management programs described in item b) regarding the Reactor Containment Building (concrete liner, post stressing system, anchor elements). This module will also support the Maintenance Rule related practices. (Author)

  18. Nanosecond Tm:Y2O3 ceramic laser passively Q-switched by a Ho:LuAG ceramic

    Science.gov (United States)

    Wang, Hui; Huang, Haitao; Wang, Shiqiang; Shen, Deyuan

    2018-02-01

    A passively Q-switched 2.05-μm Tm:Y2O3 ceramic laser, employing Ho:LuAG ceramic as a saturable absorber, was demonstrated for the first time. Under the absorbed pump power of 20.5 W, a maximum output power of 497 mW was obtained. Pulses with a minimum pulse width of 642 ns under the repetition rate of 33 kHz were achieved. Our works validate that Ho-doped materials have good potential for passive Q-switching of Tm-doped lasers at 2-μm wavelength region.

  19. Neutron activation of building materials used in the reactor shield

    International Nuclear Information System (INIS)

    Hernandez, A.T.; Perez, G.; D'Alessandro, K.

    1993-01-01

    Cuban concretes and their main components (mineral aggregates and cement) were investigated through long-lived activation products induced by neutrons from a reactor. The multielemental content in the materials studied was obtained by neutron activation analysis in an IBR-2 reactor and gamma activation analysis in an MT-25 microtron from Join Institute of Nuclear Research of Dubna. After irradiation of building materials for 30 years by a neutron flow of unitary density, induced radioactivity was calculated according to experimental data. The comparative evaluation of different concretes aggregates and two types of cement related to the activation properties is discussed

  20. Seismic analysis of a reactor building with eccentric layout

    International Nuclear Information System (INIS)

    Itoh, T.; Deng, D.Z.F.; Lui, K.

    1987-01-01

    Conventional design for a reactor building in a high seismic area has adopted an essentially concentric layout in response to fear of excessive torsional effect due to horizontal seismic load on an eccentric plant. This concentric layout requirement generally results in an inflexible arrangement of the plant facilities and thus increases the plant volume. This study is performed to investigate the effect of eccentricity on the overall seismic structural response and to provide technical information in this regard to substantiate the volume reduction of the overall power plant. The plant layout is evolved from the Bechtel standard plan of a PWR plant by integrating the reactor building and the auxiliary building into a combined building supported on a common basemat. This plant layout is optimized for volume utilization and to reduce the length of piping systems. The mass centers at various elevations of the combined building do not coincide with the rigidity center (RC) of the respective floor and the geometric center of the basemat, thus creating an eccentric response of the building in a seismic environment. Therefore, the torsional effects of the structure have to be taken into account in the seismic analysis

  1. Calculation of prefabricated part of WWR-K reactor building

    International Nuclear Information System (INIS)

    Belyashova, N.N.; Aptikaev, F.F.; Kopnichev, Yu.F.

    1998-01-01

    According of factual characteristics a strength and deformation of over-land part of carrier constructions under construction movement is defined. Direct dynamical calculation of design elements under action of inertial loads from supports shifts shows, that seismic stability of enclosing construction is not ensured. Possibly practically total collapse of coating construction is possibly, under which following levels of damages of internal design constructions of reactor central room have been forecasted: 1. Fall of destroyed design construction on reactor vessel in time moment (1.56-1.59 s) after coming to building of earthquake seismic waves of 10 balls. 2. It is possibly cracks formation in radial direction in lower part of reactor cap, but destroying of cap does not incident; 3. It is possibly cracks formation within stretched concrete zone of reactor construction at the mark from - 0.859 up to 0.100. Destroy of concrete's compressive zone of reactor construction have not being expected. 4. Collapse of reactor first contour coating constructions have not being expected

  2. Considerations on safety against seismic excitations in the project of reactor auxiliary building and control building in nuclear power plants

    International Nuclear Information System (INIS)

    Santos, S.H.C.; Castro Monteiro, I. de

    1986-01-01

    The seismic requests to be considered in the project of main buildings of a nuclear power plant are discussed. The models for global seismic analysis of nuclear power plant structures, as well as models for global strength distribution are presented. The models for analysing reactor auxiliary building and control building, which together with the reactor building and turbine building form the main energy generation complex in a nuclear power plant, are described. (M.C.K.) [pt

  3. Development on seismic isolation technique for the reactor buildings

    International Nuclear Information System (INIS)

    Yamauchi, Hiroyuki; Okawa, Izuru

    2000-01-01

    In recent earthquake, a large acceleration was observed not only at horizontal movement but also at vertical one, it is essential to grasp performance of limit state and seismic isolation mechanism at its state for actualizing a seismic isolation building. This study aims at experimentally and analytically investigating the behavior required for applying a seismic isolation construction method to a nuclear relating facility under its limit state, and at verifying seismic performance of whole of the reactor building due to seismic observation using a model building. In 1998 fiscal year, investigation on recovery force performance of high dumping laminated rubber at vertical (axial) direction, investigation on effect on behavior of basic performance and upper structure forming at seismic isolation structure on earthquake, experimental plan on grasping basic performance of seismic isolation apparatus, and seismic observation of seismic isolation building for a whole model, were carried out. (G.K.)

  4. Reactor building seismic analysis of a PWR type - NPP

    International Nuclear Information System (INIS)

    Kakubo, Masao

    1983-01-01

    Earthquake engineering studies raised up in Brazil during design licensing and construction phases of Almirante Alvaro Alberto NPP, units 1 and 2. State of art of soil - structure interaction analysis with particular reference to the impedance function calculation analysis with particular reference to the impedance function calculation of a group of pile is presented in this M.Sc. Dissertation, as an example the reactor building dynamic response of a 1325 MWe NPP PWR type is calculated. The reactor building is supported by a pile foundation with 2002 end bearing piles. Upper and lower bound soil parameters are considered in order to observe their influence on dynamic response of structure. Dynamic response distribution on pile heads show pile-soil-pile interaction effects. (author)

  5. Earthquake response of nuclear reactor buildings deeply embedded in soil

    International Nuclear Information System (INIS)

    Masao, T.; Takasaki, Y.; Hirasawa, M.; Okajima, M.; Yamamoto, S.; Kawata, E.; Koori, Y.; Ochiai, S.; Shimizu, N.

    1980-01-01

    This paper is concerned with experimental and analytical studies to investigate dynamic behavior of deeply embedded structures such as nuclear reactor buildings. The principal points studied are as follows: (1) Examination of stiffness and radiation damping effects according to embedded depth, (2) verification for distributions of earth pressure according to embedded depth, (3) differences of response characteristics during oscillation according to embedded depth, and (4) proposal of an analytical method for seismic design. Experimental studies were performed by two ways: forced vibration test, and earthquake observation against a rigid body model embedded in soil. Three analytical procedures were performed to compare experimental results and to examine the relation between each procedure. Finally, the dynamic behavior for nuclear reactor buildings with different embedded depths were evaluated by an analytical method. (orig.)

  6. Reactor

    International Nuclear Information System (INIS)

    Ikeda, Masaomi; Kashimura, Kazuo; Inoue, Kazuyuki; Nishioka, Kazuya.

    1979-01-01

    Purpose: To facilitate the construction of a reactor containment building, whereby the inspections of the outer wall of a reactor container after the completion of the construction of the reactor building can be easily carried out. Constitution: In a reactor accommodated in a container encircled by a building wall, a space is provided between the container and the building wall encircling the container, and a metal wall is provided in the space so that it is fitted in the building wall in an attachable or detatchable manner. (Aizawa, K.)

  7. Determination of the NPP Cernavoda reactor building seismic response

    International Nuclear Information System (INIS)

    Krutzik, N.J.; Rotaru, I.; Bobei, M.; Mingiuc, C.; Serban, V.

    1997-01-01

    Seismic input for systems and equipment installed in buildings depends on: - the seismic movement in free field on site; - the building movement in the soil; - the building deflection. The percentage of the 3 movements for the system and equipment input, depends on the position of the systems and equipment inside the building as well on the type of the foundation soil. The type of the foundation soil is important because if it is stiff it transfers a lot of energy to the building, energy which amplify the movement of the building on the top. If the foundation soil is soft, it accommodates the overall movement of the building in the soil, amplifying the movement to lower levels and the building response is attenuated if a resonance phenomenon between the whole building movement and the seismic excitation does not exist. This input is given with the design floor response spectra (FRS), in the logarithmic scale and seismic anchor movement (SAM). The design floor response spectra for NPP Cernavoda U1 Nuclear Building were determined in several stages starting with simple models (STICK type) without twisting movement and ending with detailed 3-dimensional models. From the point of view of dynamic behavior, the Reactor Building can be considered to be made up of 4 sub-structures: the containment building, internal structures containing separate elements such as the reactor vault, the fuel transfer structure and itself. Each sub-structure has its own movement (some of the structures present also some local effects) which combines with the overall movement of the building in the soil and the seismic excitation produce the total movement so that the response spectrum for each point of the sub-structure is specific. One should note that for structures which also show the twisting effect, the selection of the points on the floor, for the determination on the response spectra, is an engineering decision so that the response should be relevant for the equipment installed on the

  8. HELB Analysis for ESBWR Reactor Building and Main Steam Tunnel

    Energy Technology Data Exchange (ETDEWEB)

    Noguera Oliva, O.

    2011-07-01

    The Reactor Building compartments and tbe Main Steam Tunnel are modeled using GOTHIC 7.2a. These models are based on Control Volumes (Rooms/Compartments/Regions), Flow Paths (junctions such as vent path or any opening) and Boundary Conditions (Mass and energy releases and outside conditions). Due to the different break locations, four models are built to analyze the short-term pressurization response. Are shown the cases analyzed, the results obtained and the models used for this purpose.

  9. Study on reactor building structure using ultrahigh strength materials, 1

    International Nuclear Information System (INIS)

    Ishimura, Kikuo; Odajima, Masahiro; Irino, Kazuo; Hashiba, Toshio.

    1991-01-01

    This study was promoted to be aimed at realization of the optimal nuclear reactor building structure of the future. As the first step, the study regarding ultrahigh strength reinforced concrete (abbr. RC) shear wall was selected. As the result of various tests, the application of ultrahigh strength RC shear walls was verified. The tests conducted were relevant to; ultrahigh strength concrete material tests; pure shear tests of RC flat panels; and bending shear tests and its simulation analysis of RC shear walls. (author)

  10. Thermal Reactor Safety

    Energy Technology Data Exchange (ETDEWEB)

    1980-06-01

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

  11. Thermal Reactor Safety

    International Nuclear Information System (INIS)

    1980-06-01

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

  12. Response characteristics of reactor building on weathered soft rock ground

    International Nuclear Information System (INIS)

    Hirata, Kazuta; Tochigi, Hitoshi

    1991-01-01

    The purpose of this study is to investigate the seismic stability of nuclear power plants on layered soft bedrock grounds, focusing on the seismic response of reactor buildings. In this case, the soft bedrock grounds refer to the weathered soft bedrocks with several tens meter thickness overlaying hard bedrocks. Under this condition, there are two subjects regarding the estimation of the seismic response of reactor buildings. One is the estimation of the seismic response of surface ground, and another is the estimation of soil-structure interaction characteristics for the structures embedded in the layered grounds with low impedandce ratio between the surface ground and the bedrock. Paying attention to these subjects, many cases of seismic response analysis were carried out, and the following facts were clarified. In the soft rock grounds overlaying hard bedrocks, it was proved that the response acceleration was larger than the case of uniform hard bedrocks. A simplified sway and rocking model was proposed to consider soil-structure interaction. It was proved that the response of reactor buildings was small when the effect of embedment was considered. (K.I.)

  13. Study of vibration analysis for nuclear reactor building

    International Nuclear Information System (INIS)

    Hirashima, Shin-ichi

    1978-01-01

    The mutual interference between the contiguous buildings with separate foundations and also that between the outer wall under the ground and the foundation bottom of the building were taken into consideration for the vibration analysis with spring-mass system. For two contiguous foundations of buildings it was attempted to represent the static mutual interference by a spring-mass system model. The theoretical analysis formulas are shown for the combination of the vertical movement and rocking motion, and for the interfering forces between the foundation and the outer wall of a building. The method of extending the model to dynamic one is explained. Several spring constants utilized in the analysis were obtained, for example, for mutual interference springs regarding vertical motion, mutual interfering springs for the foundation and the outer wall of a building and the mutual interference springs concerning horizontal movement. These models and analysis were applied to the BWR-MARK II-1100 MW nuclear reactor building and the contiguous turbine building. The structures and level relations of two buildings are shown, and the spring-mass system model for these buildings is expressed. The masses of about 20, the weights, the rotating inertia, the sectional moment of inertia, the spring constant and the damping coefficient for each mass are tabulated. As the results, the peak displacements occur at 2.556 Hz, 6.918 Hz, 10.43 Hz and 13.85 Hz. The damping coefficient is large and about 10 - 30% at the lower order modes. The calculated and the measured vibration characteristics for the BWR plant buildings are not much different, and this spring-mass system model is verified to be adequate. (Nakai, Y.)

  14. Dynamic analysis of a reactor building on alluvial soil

    International Nuclear Information System (INIS)

    Arya, A.S.; Chandrasekaran, A.R.; Paul, D.K.; Warudkar, A.S.

    1977-01-01

    The reactor building consists of reinforced concrete internal framed structure enclosed in double containment shells of prestressed and reinforced concrete all resting on a common massive raft. The external cylindrical shell is capped by a spherical dome while the internal shell carries a cellular gird slab. The building is partially buried under ground. The soil consists of alluvial going to 1000 m depth. The site lies in a moderate seismic zone. The paper presents the dynamic analysis of the building including soil-structure interaction. The mathematical model consists of four parallel, suitably interconnected struxtures, namely inner containment, outer containment, internal frame and the calandria vault. Each one of the parallel structures consists of lumped-mass beam elements. The soil below the raft and on the sides of outer containment shell is represented by elastic springs in both horizontal and vertical directions. The various assumpions required to be made in developing the mathematical model are briefly discussed in the paper. (Auth.)

  15. Seismic resistant analysis of coupled model of reactor coolant system and reactor building

    International Nuclear Information System (INIS)

    Wang Xiaowen; Xia Zufeng

    2005-01-01

    Reactor coolant system(RCS) and reactor building are actually coupled with each other. SRP (Revision 2) edited by USNRC particularly pointed out in 3.7.2 that RCS, which is considered a subsystem but is usually analyzed using a coupled model with building. Under this background, this paper selects PC-NPP as a study object, and seismic resistant analysis is performed with a coupled model of building and RCS using response spectrum method and time history method. Finally, analyzed results are compared with those of uncoupled RCS model. In the analysis, building is simulated with cantilever beam model of shear wall combination. In the uncoupled model, each supporting of equipment is modeled using elastic beam element with actual supporting stiffness, which is connected to a rigid cantilever (single-point input) and to an elastic cantilever (multipoint input). Seismic load of coupled model is input from the bottom of building. After comparison, it is shown that the effect of interaction between RCS and building can not be ignored, and the uncoupled model for seismic resistant analysis is inappropriate to be applied in actual seismic design. Through this research, we can control the seismic analysis technique in coupled model and enhance our analysis level of NPP. (authors)

  16. Structural design of SBWR reactor building complex using microcomputers

    International Nuclear Information System (INIS)

    Mandagi, K.; Rajagopal, R.S.; Sawhney, P.S.; Gou, P.F.

    1993-01-01

    The design concept of Simplified Boiling Water Reactor (SBWR) plant is based on simplicity and passive features to enhance safety and reliability, improve performance, and increase economic viability. The SBWR utilizes passive systems such as Gravity Driven Core-Cooling System (GDCS) and Passive Containment Cooling System (PCCS). To suit these design features the Reactor Building (RB) complex of the SBWR is configured as an integrated structure consisting of a cylindrical Reinforced Concrete Containment Vessel (RCCV) surrounded by square reinforced concrete safety envelope and outer box structures, all sharing a common reinforced concrete basemat. This paper describes the structural analysis and design aspects of the RB complex. A 3D STARDYNE finite element model has been developed for the structural analysis of the complex using a PC Compaq 486/33L microcomputer. The structural analysis is performed for service and factored load conditions for the applicable loading combinations. The dynamic responses of containment structures due to pool hydrodynamic loads have been calculated by an axisymmetric shell model using COSMOS/M program. The RCCV is designed in accordance with ASME Section 3, Division 2 Code. The rest of the RB which is classified as Seismic Category 1 structure is designed in accordance with the ACI 349 Code. This paper shows that microcomputers can be efficiently used for the analysis and design of large and complex structures such as RCCV and Reactor Building complex. The use of microcomputers can result in significant savings in the computational cost compared with that of mainframe computers

  17. Status of the TMI-2 core: a review of damage assessments

    International Nuclear Information System (INIS)

    Croucher, D.W.

    1981-01-01

    Assessments of the damage within the core of the Three Mile Island Unit 2 reactor, performed by reconstructing the transient thermal-hydraulic sequence of events, estimating the amount of hydrogen generation, and evaluating the amount of fission products released, are reviewed and summarized. Minimum and maximum bounds of damage to the core are identified

  18. Building reactor operator sustain expert system with C language integrated production system

    International Nuclear Information System (INIS)

    Ouyang Qin; Hu Shouyin; Wang Ruipian

    2002-01-01

    The development of the reactor operator sustain expert system is introduced, the capability of building reactor operator sustain expert system is discussed with C Language Integrated Production System (Clips), and a simple antitype of expert system is illustrated. The limitation of building reactor operator sustain expert system with Clips is also discussed

  19. Forced vibration test of BWR type nuclear reactor buildings considering through-soil coupling between adjacent buildings

    International Nuclear Information System (INIS)

    Mizuno, Norihiro; Moribe, Isamu; Sugiyama, Nobuo; Tsushima, Yoshiyuki; Kushida, Hiroshi

    1980-01-01

    The forced vibration test on two adjacent BWR type reactor buildings in the Hamaoka Nuclear Power Station, Chubu Electric Power Co., Inc., was carried out in 1977. The experiment on No. 1 reactor building had been made in 1973 and 1974 to evaluate the dissipation of vibrational energy to the earth. In 1977, No. 2 reactor building was under construction and nearly completed adjacently to the No. 1 reactor building, in which the reactor has been in operation since 1975. Three large vibration exciters with maximum exciting force of 150 tons each were installed on the 5th floor of the No. 2 reactor building, and the vibration ranging from 1 to 20 Hz was applied to the building in north-south, east-west and vertical directions. For obtaining the responses of the buildings, 24 and 33 pickups were placed on the floors of No. 1 and No. 2 buildings, respectively, and 17 dynamic earth pressure cells buried under the foundation of the No. 2 building were also employed. The natural frequency, viscous damping ratio and natural mode were analyzed by the regression analysis of many degrees of freedom, using the resonance curves with complex forms and many peaks. The vibrational behaviors of two adjacent reactor buildings coupled through the earth were given as the result of experiment. (J.P.N.)

  20. Evaluation for rigidity of box construction of nuclear reactor building

    International Nuclear Information System (INIS)

    Yamakawa, Tetsuo

    1979-01-01

    A huge box-shaped structure (hereafter, called box construction) of reinforced concrete is presently utilized as the reactor building structure in nuclear power plants. Evaluation of the rigidity of the huge box construction is required for making a vibration analysis model of nuclear reactor buildings. It is necessary to handle the box construction as the plates to which the force in plane is applied. This paper describes that the bending theory in elementary beam theory is equivalent to a peculiar, orthogonally anisotropic plate, the shearing rigidity and film rigidity in y direction of which are put to infinity and the Poisson's ratio is put to zero, viewed from the two-dimensional theory of elasticity. The form factor of 1.2 for shearing deformation in rectangular cross section was calculated from the parabolic distribution of shearing stress intensity, and it is the maximum value. The factor is equal to 1.2 for slender beams, but smaller than 1.2 for short and thick beams, having tendency to converge to 1.0. The non-conformity of boundary conditions regarding the shearing force at the both ends of cantilevers does not affect very seriously the evaluation of shearing rigidity. From the above results, it was found that the application of the theory to the box construction was able to give the rigidity evaluation with sufficient engineering accuracy. The theory can also be applied to the evaluation of tube type ultrahigh buildings. (Wakatsuki, Y.)

  1. Structure of steel reactor building and construction method therefor

    International Nuclear Information System (INIS)

    Yamakawa, Toshikimi.

    1997-01-01

    The building of the present invention contains a reactor pressure vessel, and has double steel plate walls endurable to elevation of inner pressure and keeping airtightness, and shielding concretes are filled between the double steel plate walls. It also has empty double steel plate walls not filled with concretes and has pipelines, vent ducts, wirings and a support structures for attaching them between the double steel plate walls. It is endurable to a great inner pressure satisfactory and keeps airtightness by the two spaced steel plates. It can be greatly reduced in the weight, and can be manufactured efficiently with high quality in a plant by so called module construction, and the dimension of the entire of the reactor building can be reduced. It is constructed in a dock, transported on the sea while having the space between the two steel plate walls as a ballast tanks, placed in the site, and shielding concretes are filled between the double steel plate walls. The term for the construction can be reduced, and the cost for the construction can be saved. (N.H.)

  2. Culham conceptual Tokamak reactor MkII. Conceptual layout of buildings for a twin reactor power station

    International Nuclear Information System (INIS)

    Guthrie, J.A.S.; Harding, N.H.

    1981-01-01

    This paper discusses the conceptual design of the nuclear complex of a 2400 MWe twin fusion reactor power station utilising common services and a single containment building. The design is based upon environmental and maintenance logistical requirements, the provision of adequate storage, workshop and construction facilities and the constraints imposed by the geometry of the main and auxiliary reactor coolant systems. (author)

  3. Turning points in reactor design

    Energy Technology Data Exchange (ETDEWEB)

    Beckjord, E.S.

    1995-09-01

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

  4. Turning points in reactor design

    International Nuclear Information System (INIS)

    Beckjord, E.S.

    1995-01-01

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

  5. Seismic response of a reactor building on rocky foundation

    International Nuclear Information System (INIS)

    Thakkar, S.K.; Tilak, M.M.; Verma, U.S.P.; Ramachandran, V.

    1987-01-01

    The seismic response of a reactor building on a rocky foundation has been studied by the finite element model. The influence of base fixity, inertia of foundation, stiffness of internal structure and foundation rock properties is studied on the seismic response. The conclusions of this study are summarised as follows: the response due to one mode alone is found to be significant in the case of rocky foundation; the foundation - structure interaction has relatively small contribution to the total response; the consideration of inertia of foundation may lead to very high seismic respones. More realistic method of considering inertia effect of foundation is therefore necessary; the increase in the elasticity of foundation by one hundred percent does not cause variations in vertical stresses in the outer and inner shells by more than 10 percent

  6. Monitoring actual temperatures in Susquehanna SES reactor buildings

    International Nuclear Information System (INIS)

    Derkacs, A.P.

    1991-01-01

    PP and L has been monitoring temperatures in the Susquehanna SES reactor building with digital temperature recorders since 1986. In early 1990, data from four representative areas was analyzed to determine the temperature in each area which would produce the same rate of degradation as the distribution of actual temperatures recorded over about 40 months. From these effective average temperatures, qualified life multipliers were determined for activation energies in the range of 0.5 to 1.5 and those multipliers were used to estimate new qualified lives and the number of replacements which might be saved during the life of the plant. The results indicate that pursuing a program of determining EQ qualified lives from actual temperatures, rather than maximum design basis temperatures, will provide a substantial payback in reduced EQ driven maintenance

  7. Dynamic analysis of a reactor building on alluvial soil

    International Nuclear Information System (INIS)

    Arya, A.S.; Chandrasekaran, A.R.; Paul, D.K.

    1977-01-01

    The reactor building consists of reinforced concrete internal framed structure enclosed in double containment shells of prestressed and reinforced concrete all resting on a common massive raft. The external cylindrical shell is capped by a spherical dome while the internal shell carries a cellular grid slab. The building is partially buried under ground. The soil consists of alluvial going to 1000 m depth. The site lies in a moderate seismic zone. The paper presents the dynamic analysis of the building including soil-structure interaction. The mathematical model consists of four parallel, suitably interconnected structures, namely inner containment, outer containment, internal frame and the calandria vault. Each one of the parallel structures consists of lumped-mass beam elements. The soil below the raft and on the sides of outer containment shell is represented by elastic springs in both horizontal and vertical directions. The various assumptions required to be made in developing the mathematical model are briefly discussed in the paper. Transfer matrix technique has been used to determine the frequencies and mode shapes. The deformations due to bending, shear and effect of the rotary inertia have been included. Various alternatives of laterally interconnecting the internals and the shells have been examined and the best alternative from earthquake considerations has been obtained. In the study, the effect of internal structure flexibility and Calandria vault flexibility on the whole building have been studied. The resulting base raft motion and the structural timewise response of all floors have been determined for the design basis (safe shutdown) earthquake by mode superposition

  8. QA in the design and fabrication of the TMI-2 rail cask

    International Nuclear Information System (INIS)

    Hayes, G.R.

    1988-01-01

    EGandG Idaho, Inc., acting on behalf of the US Department of Energy, is responsible for transporting core debris from Three Mile Island-Unit 2 to the Idaho National Engineering Laboratory. Transportation of the debris is being accomplished using an NRC licensed container, called the NuPac 125-B. This paper describes the NuPac 125-B Rail Cask and the quality assurance (QA) requirements for that system. Also discussed are the QA roles of the various organizations involved in designing, building, inspecting and testing the NuPac 125-B. The paper presents QA/QC systems implemented during the design, procurement, and fabrication of the cask to assure compliance with all applicable technical codes, standards and regulations. It also goes beyond the requirements aspect and describes unique QA/QC measures employed to assure that the cask was built with minimum QA problems. Finally, the lessons learned from the NuPac 125-B project is discussed. 4 refs., 4 figs

  9. Cassette blanket and vacuum building: key elements in fusion reactor maintenance

    International Nuclear Information System (INIS)

    Werner, R.W.

    1977-01-01

    The integration of two concepts important to fusion power reactors is discussed. The first concept is the vacuum building which improves upon the current fusion reactor designs. The second concept, the use of the cassette blanket within the vacuum building environment, introduces four major improvements in blanket design: cassette blanket module, zoning concept, rectangular blanket concept, and internal tritium recovery

  10. Reactor building design of nuclear power plant ATUCHA II, Argentina

    International Nuclear Information System (INIS)

    Rufino, R.E.; Hermann, E.R.; Richter, E.

    1984-01-01

    It is presented the civil engineering project carried out by the joint venture Hochtief - Techint-Bignoli (HTB) for the reactor building at the Atucha II power plant (PHWR of 745 MWe) in Buenos Aires. All the other civil projects at Atucha II are also being carried out by HTB. This building has the same general characteristics of the PWR plants developed by KWU in Germany, known for the spherical steel containment 56m in diameter. Nevertheless, it differs from those principally in the equipment lay-out and the remarkable foundation depth. From the basic engineering provided by ENACE, the joint venture has had to face the challenge of designing a tridimensional structure of large size. This has necessitated using simplified models which had to be superimposed, since the use of only one spatial mode would be highly inadequate, lacking the flexibility necessary to absorb the numerous modifications that this type of project undergoes during construction. In addition, this procedure has eliminated resorting to numerous and costly computer processings. (Author) [pt

  11. Assessment of chemical processes for the post-accident decontamination of reactor-coolant systems. Final report

    International Nuclear Information System (INIS)

    Munson, L.F.; Card, C.J.; Divine, J.R.

    1983-02-01

    Previously used chemical decontamination processes and potentially useful new decontamination processes were examined for the usefulness following a reactor accident. Both generic fuel damage accidents and the accident at TMI-2 were considered. A total of fourteen processes were evaluated. Process evaluation included data in the following categories: technical description of the process, recorded past usage, effectiveness, process limitation, safety consideration, and waste management. These data were evaluated, and cost considerations were presented along with a description of the applicability of the process to TMI-2 and development and demonstration needs. Specific recommendations regarding a primary-system decontamination development program to support TMI-2 recovery were also presented

  12. Seismic simulation analysis of nuclear reactor building by soil-building interaction model

    International Nuclear Information System (INIS)

    Muto, K.; Kobayashi, T.; Motohashi, S.; Kusano, N.; Mizuno, N.; Sugiyama, N.

    1981-01-01

    Seismic simulation analysis were performed for evaluating soil-structure interaction effects by an analytical approach using a 'Lattice Model' developed by the authors. The purpose of this paper is to check the adequacy of this procedure for analyzing soil-structure interaction by means of comparing computed results with recorded ones. The 'Lattice Model' approach employs a lumped mass interactive model, in which not only the structure but also the underlying and/or surrounding soil are modeled as descretized elements. The analytical model used for this study extends about 310 m in the horizontal direction and about 103 m in depth. The reactor building is modeled as three shearing-bending sticks (outer wall, inner wall and shield wall) and the underlying and surrounding soil are divided into four shearing sticks (column directly beneath the reactor building, adjacent, near and distant columns). A corresponding input base motion for the 'Lattice Model' was determined by a deconvolution analysis using a recorded motion at elevation -18.5 m in the free-field. The results of this simulation analysis were shown to be in reasonably good agreement with the recorded ones in the forms of the distribution of ground motions and structural responses, acceleration time histories and related response spectra. These results showed that the 'Lattice Model' approach was an appropriate one to estimate the soil-structure interaction effects. (orig./HP)

  13. The TMI-2 accident

    International Nuclear Information System (INIS)

    Loureiro, L.A.

    1986-01-01

    A critical study about the technical and man-related facts in order to establish what is considered the worst commercial nuclear power accident until 1986. Radiological consequences and stress to the public are considered in contrast to antinuclear groups. This descriptive and technical study has the purpose to document written and oral opinions obtained abroad and then explain to the public in an easy language terminology. Preliminary study describing safety related systems fails and the accident itself with minute to minute description, conduct to the consequences and then, to learned lessons

  14. Isolator arrangement design and 3D solid modeling for KALIMER reactor building

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Jae Han; Yoo, Bong; Koo, G. H. [Korea Atomic Energy Research Institute, Taejeon (Korea)

    2000-03-01

    The axial loads on the isolators under upper basemat of KALIMER reactor building vary according to the weight distribution of reactor building, so the unbalanced deflection and the stresses can be occurred on upper basemat supported by isolators. For evaluating the phenomena, the axial forces on the isolators are calculated using the finite element model of reactor building and isolators. Optimal design study of isolator arrangement is performed, and the structural integrity of isolators and upper basemat is evaluated. For the concept design of KALIMER building, 3-dimensional solid models using IDEAS program are built and some design modifications for reactor building are implemented for the latest design. 11 refs., 19 figs., 7 tabs. (Author)

  15. Effects of soil stiffness and embedment on reactor building response

    International Nuclear Information System (INIS)

    Michalopoulos, A.P.; Vardanega, C.; Cornaggia, L.

    1981-01-01

    A parametric study was made to assess the influence of soil conditions and foundation embedment depth on the floor response spectra for a reactor building. The analyses incorporated soft, medium and hard soils, and three different embedment depths, in a seismic environment described by a 0.36 g peak ground acceleration. The shear wave velocity profiles for the soft, medium and hard soil conditions, were assumed to increase in proportion to the square root of depth from their ground surface values of 300, 600 and 900 meters per second, respectively. Foundation embedment depths of zero, eight and fourteen meters were analyzed using elastic half-space theory, accounting for kinematic interaction. The variation of shear modulus with depth under earthquake excitation was determined using a deconvolution process. Horizontal and vertical synthetic time histories, matching the USNRC Regulatory Guide 1.60 design ground response spectra, were applied at the ground surface and then deconvolved to the foundation level to obtain the input for the soil-structure model. The mathematical model of the superstructure consisted of four lumped-mass close-coupled systems, representing containment shells and components, while the foundation mat was modeled as rigid. Lumped soil compliances (springs and dashpots) were used to represent the horizontal, vertical and rotational modes of vibration. The dynamic analyses were performed utilizing the computer code DAPSYS, and consisted of mode frequency analyses and modal superposition. Modal damping was computed as a weighted average of structural and soil (radiation and material) damping, using the strain energy stored in the respective components as the weighting factor and distinguishing the hysteric nature of the structural and soil material damping, and the viscous nature of the soil radiation damping. (orig./RW)

  16. Prediction of hydrogen distribution in the reactor building in CANDU6 plant

    International Nuclear Information System (INIS)

    Jin, Y.; Song, Y.

    2008-01-01

    The CANDU plants have a lot of zircaloy. The fuel cladding, calandria tubes and pressure tubes are made of zircaloy. The zircaloy can be oxidized and hydrogen is generated during severe accident progression. The detonation or deflagration to detonation transition (DDT) due to hydrogen combustion may occur if the local hydrogen concentration or global hydrogen concentration exceeds certain value. The detonation may result in the rupture of the reactor building. The inside of the reactor building of CANDU plants is complex. So prediction of hydrogen distribution in the reactor building is important. This prediction is made using ISAAC code and GOTHIC code. ISAAC code partitioned the reactor building in to 7 compartments. GOTHIC code modeled the CANDU6 reactor building using 12 nodes. The hydrogen concentrations in the various compartments in the reactor building are compared. GOTHIC code slightly underpredicts hydrogen concentration in the F/M rooms than ISAAC code, but trend is same. The hydrogen concentration in the boiler room and the moderator room shows almost same as for both codes. (author)

  17. Dynamic analysis of reactor containment building using axisymmetric finite element model

    International Nuclear Information System (INIS)

    Thakkar, S.K.; Dubey, R.N.

    1989-01-01

    The structural safety of nuclear reactor building during earthquake is of great importance in view of possibility of radiation hazards. The rational evaluation of forces and displacements in various portions of structure and foundation during strong ground motion is most important for safe performance and economic design of the reactor building. The accuracy of results of dynamic analysis is naturally dependent on the type of mathematical model employed. Three types of mathematical models are employed for dynamic analysis of reactor building beam model axisymmetric finite element model and three dimensional model. In this paper emphasis is laid on axisymmetric model. This model of containment building is considered a reinfinement over conventional beam model of the structure. The nuclear reactor building on a rocky foundation is considered herein. The foundation-structure interaction is relatively less in this condition. The objective of the paper is to highlight the significance of modelling of non-axisymmetric portion of building, such as reactor internals by equivalent axisymmetric body, on the structural response of the building

  18. Method of decommissioning nuclear reactor building by utilizing sea water buyoancy

    International Nuclear Information System (INIS)

    Iwashima, Sumio; Ogoshi, Shigeru; Kobari, Shin-ichi.

    1989-01-01

    Upon dismantling nuclear reactor buildings, peripheral yards are excavated and channels leading to sea shore are formed. Since the outer walls of the reactor buildings are made of iron-reinforced concretes, the opening poritons are grouted with concretes to attain a tightly such closed structure that radioactive wastes, etc. in the inside are not flown out upon reactor discommisioning. Peripheral buildings at relatively low level of radiation contaminations are dismantled and withdrawn. The fundations of the nuclear reactor buildings were dug out and jacked to separate base rocks and the reactor buildings. Then, sea water is introduced into the water channels to entirely float up the buildings. A water gate is disposed in the water channel on the side of sea shore to control the level of sea water. The buildings are moved and guided to the sea shore and towed to a site optimum as a permanent storage area and then burried in that place. The operation period for the discommissioning work can greatly be shortened and the radiation dose and the amount of the wastes can be reduced. (T.M.)

  19. Nonlinear rocking analysis of nuclear reactor building simultaneously subjected to horizontal and vertical earthquake motions

    International Nuclear Information System (INIS)

    Muto, K.; Kobayashi, T.; Motohashi, S.; Mizuno, N.; Moribe, I.; Sugiyama, N.; Suzuki, T.

    1983-01-01

    In the dynamic analysis of a reactor building, the response acceleration of the building is largely amplified during an earthquake and it's base mat may be lifted by a response overturning moment caused by a strong earthquake. And it makes geometrical nonlinear interaction between the base mat and rock foundation beneath it, which produces very complex phenomena by horizontal and vertical earthquake motions. In this paper, the dynamic behavior of a BWR type reactor building is studied considering the uplift of the base mat subjected to simultaneous horizontal and vertical earthquake motions. Results: (1) The horizontal maximum response values of the building (acceleration, shear force and bending moment) have little effect by the existence of the vertical input. (2) The vertical acceleration of the building is generated by the uplift of the base mat caused by horizontal input. (3) The horizontal acceleration response spectrum of the building takes some effect by the uplift within it's high frequency range. (orig./HP)

  20. Aging management program of the reactor building concrete at Point Lepreau Generating Station

    Directory of Open Access Journals (Sweden)

    Gendron T.

    2011-04-01

    Full Text Available In order for New Brunswick Power Nuclear (NBPN to control the risks of degradation of the concrete reactor building at the Point Lepreau Generating Station (PLGS the development of an aging management plan (AMP was initiated. The intention of this plan was to determine the requirements for specific structural components of concrete of the reactor building that require regular inspection and maintenance to ensure the safe and reliable operation of the plant. The document is currently in draft form and presents an integrated methodology for the application of an AMP for the concrete of the reactor building. The current AMP addresses the reactor building structure and various components, such as joint sealant and liners that are integral to the structure. It does not include internal components housed within the structure. This paper provides background information regarding the document developed and the strategy developed to manage potential degradation of the concrete of the reactor building, as well as specific programs and preventive and corrective maintenance activities initiated.

  1. Building a dynamic code to simulate new reactor concepts

    International Nuclear Information System (INIS)

    Catsaros, N.; Gaveau, B.; Jaekel, M.-T.; Maillard, J.; Maurel, G.; Savva, P.; Silva, J.; Varvayanni, M.

    2012-01-01

    Highlights: ► We develop a stochastic neutronic code based on an existing High Energy Physics code. ► The code simulates innovative reactor designs including Accelerator Driven Systems. ► Core materials evolution will be dynamically simulated, including fuel burnup. ► Continuous feedback between the main inter-related parameters will be established. ► A description of the current research development and achievements is also given. - Abstract: Innovative nuclear reactor designs have been proposed, such as the Accelerator Driven Systems (ADSs), the “candle” reactors, etc. These reactor designs introduce computational nuclear technology problems the solution of which necessitates a new, global and dynamic computational approach of the system. A continuous feedback procedure must be established between the main inter-related parameters of the system such as the chemical, physical and isotopic composition of the core, the neutron flux distribution and the temperature field. Furthermore, as far as ADSs are concerned, the ability of the computational tool to simulate the nuclear cascade created from the interaction of accelerated protons with the spallation target as well as the produced neutrons, is also required. The new Monte Carlo code ANET (Advanced Neutronics with Evolution and Thermal hydraulic feedback) is being developed based on the GEANT3 High Energy Physics code, aiming to progressively satisfy all the above requirements. A description of the capabilities and methodologies implemented in the present version of ANET is given here, together with some illustrative applications of the code.

  2. Study on vertical seismic response model of BWR-type reactor building

    International Nuclear Information System (INIS)

    Konno, T.; Motohashi, S.; Izumi, M.; Iizuka, S.

    1993-01-01

    A study on advanced seismic design for LWR has been carried out by the Nuclear Power Engineering Corporation (NUPEC), under the sponsorship of the Ministry of International Trade and Industry (MITI) of Japan. As a part of the study, it has been investigated to construct an accurate analytical model of reactor buildings for a seismic response analysis, which can reasonably represent dynamic characteristics of the building. In Japan, vibration models of reactor buildings for horizontal ground motion have been studied and examined through many simulation analyses for forced vibration tests and earthquake observations of actual buildings. And now it is possible to establish a reliable horizontal vibration model on the basis of multi-lumped mass and spring model. However, vertical vibration models have not been so much studied as horizontal models, due to less observed data for vertical motions. In this paper, the vertical seismic response models of a BWR-type reactor building including soil-structure interaction effect are numerically studied, by comparing the dynamic characteristics of (1) three dimensional finite element model, (2) multi-stick lumped mass model with a flexible base-mat, (3) multi-stick lumped mass model with a rigid base-mat and (4) single-stick lumped mass model. In particular, the BWR-type reactor building has the long span truss roof which is considered to be one of the critical members to vertical excitation. The modelings of the roof trusses are also studied

  3. Nonlinear seismic response analysis of an embedded reactor building based on the substructure approach

    International Nuclear Information System (INIS)

    Hasegawa, M.; Ichikawa, T.; Nakai, S.; Watanabe, T.

    1987-01-01

    A practical method to calculate the elasto-plastic seismic response of structures considering the dynamic soil-structure interaction is presented. The substructure technique in the time domain is utilized in the proposed method. A simple soil spring system with the coupling effects which are usually evaluated by the impedance matrix is introduced to consider the soil-structure interaction for embedded structures. As a numerical example, the response of a BWR-MARK II type reactor building embedded in the layered soil is calculated. The accuracy of the present method is verified by comparing its numerical results with exact solutions. The nonlinear behaivor and the soil-structure interaction effects on the response of the reactor building are also discussed in detail. It is concluded that the present method is effective for the aseismic design considering both the material nonlinearity of the nuclear reactor building and the dynamic soil-structure interaction. (orig.)

  4. Effects of non-uniform embedments on earthquake responses of nuclear reactor building

    International Nuclear Information System (INIS)

    Koyanagi, Y.; Okamoto, S.; Yoshida, K.; Inove, H.

    1989-01-01

    The nuclear reactor buildings have the portion embedded in soil. In the seismic design of such structures, it is essential to consider the effects of the embedment on the earthquake response. Most studies on these effects, however, assume the uniform embedment, i.e. the depth of the embedment is constant, which is convenient for the design and analysis. The behavior of the earthquake response considering the three-dimensional aspects of non-uniform embedment has not been made clear yet. In this paper, the authors evaluate the effects of the non-uniform embedment in an inclined ground surface on the earthquake response of a nuclear reactor building as illustrated. A typical PWR type reactor building is chosen as an analysis structure model. Four different types of embedment are set up for the comparison study. The three-dimensional analysis is carried out considering the geometry of embedment

  5. Reactor building indoor wireless network channel quality estimation using RSSI measurement of wireless sensor network

    International Nuclear Information System (INIS)

    Merat, S.

    2008-01-01

    Expanding wireless communication network reception inside reactor buildings (RB) and service wings (SW) has always been a technical challenge for operations service team. This is driven by the volume of metal equipment inside the Reactor Buildings (RB) that blocks and somehow shields the signal throughout the link. In this study, to improve wireless reception inside the Reactor Building (RB), an experimental model using indoor localization mesh based on IEEE 802.15 is developed to implement a wireless sensor network. This experimental model estimates the distance between different nodes by measuring the RSSI (Received Signal Strength Indicator). Then by using triangulation and RSSI measurement, the validity of the estimation techniques is verified to simulate the physical environmental obstacles, which block the signal transmission. (author)

  6. The Hanford Site N Reactor buildings task identification and evaluation of historic properties

    International Nuclear Information System (INIS)

    Stapp, D.C.; Marceau, T.E.

    1996-01-01

    The New Production Reactor complex at Hanford (hereafter referred to as N Reactor) is proposed to be deactivated, decommissioned, and demolished in the coming years. Recognizing that the Hanford Site has been important to the nation, state, and local community, a task was funded to examine the effects that these activities may have on the historic properties of N Reactor. The objectives of the N Reactor buildings task were to identify potential historic properties at N Reactor, to complete Historic Property Inventory forms for all structures considered eligible and ineligible for listing in the National Register of Historic Places, and to prepare a Memorandum of Agreement that identifies the measures required to mitigate any adverse effects

  7. Development of remote decontamination technologies improving internal environment of reactor buildings at Fukushima Daiichi Nuclear Power Station

    International Nuclear Information System (INIS)

    Hotta, Koji; Hayashi, Hirotada; Sakai, Hitoshi

    2016-01-01

    The reactor buildings at the Fukushima Daiichi Nuclear Power Station of Tokyo Electric Power Co., Inc., which was seriously damaged by the Great East Japan Earthquake of March 11, 2011, have been highly contaminated by radioactive materials. To safely and efficiently advance the processes related to the forthcoming decommissioning of the reactors, it is necessary to improve the hazardous environment inside the reactor buildings. During the more than four years that have elapsed since the Great East Japan Earthquake, Toshiba has been implementing various measures to reduce the ambient dose rates inside the reactor buildings through decontamination work and participation in a national project for the development of remote decontamination technologies for reactor buildings. A variety of vehicles and technologies to support decontamination work have been developed through these activities, and are significantly contributing to improvement of the environment inside the reactor buildings. (author)

  8. Reactor building integrity testing: A novel approach at Gentilly 2 - principles and methodology

    International Nuclear Information System (INIS)

    Collins, N.; Lafreniere, P.

    1991-01-01

    In 1987, Hydro-Quebec embarked on an ambitious development program to provide the Gentilly 2 nuclear power station with an effective, yet practical reactor building Integrity Test. The Gentilly 2 Integrity Test employs an innovative approach based on the reference volume concept. It is identified as the Temperature Compensation Method (TCM) System. This configuration has been demonstrated at both high and low test pressure and has achieved extraordinary precision in the leak rate measurement. The Gentilly 2 design allows the Integrity Test to be performed at a nominal 3 kPa(g) test pressure during an (11) hour period with the reactor at full power. The reactor building Pressure Test by comparison, is typically performed at high pressure 124 kPa(g)) in a 7 day window during an annual outage. The Integrity Test was developed with the goal of demonstrating containment availability. Specifically it was purported to detect a leak or hole in the 'bottled-up' reactor building greater in magnitude than an equivalent pipe of 25 mm diameter. However it is considered feasible that the high precision of the Gentilly 2 TCM System Integrity Test and a stable reactor building leak characteristic will constitute sufficient grounds for the reduction of the Pressure Test frequency. It is noted that only the TCM System has, to this date, allowed a relevant determination of the reactor building leak rate at a nominal test pressure of 3 kPa(g). Classical method tests at low pressure have lead to inconclusive results due to the high lack of precision

  9. Model tests and numerical analysis on restoring force characteristics of reactor buildings

    International Nuclear Information System (INIS)

    Uchiyama, Y.; Suzuki, S.; Akino, K.

    1987-01-01

    Seismic shear walls of nuclear reactor buildings are composed of cylindrical, truncated cone-shape, box-shape, irregular polygonal walls or its combination and they are generally heavily reinforced concrete (RC) walls. So the elasto-plastic behaviors of those RC structures in ultimate regions have many unsolved and may be considered as especially important factors for explaining nonlinear response of nuclear reactor buildings. Following these research demands, the authors have prepared a nonlinear F.E.M. code called ''SANREF'' and made an extensive study for the restoring force characteristics of the inner concrete structures (I/C) of a PWR-type containment vessel and the principal seismic shear walls of a BWR-type reactor building by some series of reduced model tests and simulation analysis for the tests results. The detailed objectives of this study can be summarized as follows: (1) Examine the effectiveness of the configurations of shear walls, reinforcement ratios, shear span ratios (M/Qd) and vertical axial stress by ''partial model test'' which simulates some independent shear walls of the PWR-type and BWR-type reactor buildings. (2) Obtain fundamental data of restoring force characteristics of the complex shaped RC structures by ''composite model test'' which models are composed of the partial model test specimens. (3) Verify the applicability of analytical methods and constitutive modelings in SANREF code for complex shaped RC structures through nonlinear simulation analysis for the composite model test

  10. Experimental study on earthquake resistant components for BWR type reactor building

    International Nuclear Information System (INIS)

    Miyazaki, Y.; Hayama, S.; Makita, T.; Tomii, Y.; Kubo, T.

    1987-01-01

    Seismic loading experiments were carried out for three types of seismic shear wall components. A set of fundamental data were obtained through the experimental studies. A maximium use of these data can be made in next stages to evaluate stiffnesses, strength, and so on of the complicated structural systems of a reactor building as well as its load-deflection characteristics. (orig./HP)

  11. Dynamic response of aircraft impact of a reactor building with protective shell on independent foundation

    International Nuclear Information System (INIS)

    Constantopoulos, I.V.; Vardanega, C.; Attalla, I.

    1981-01-01

    Aircraft impact loading can penalize significantly the design of the equipment in a conventional containment building. An alternative scheme was developed in an attempt to reduce the aircraft impact response. A preliminary study was carried out to investigate the feasibility of the alternative scheme. This study was made in such perspective and for the purpose of comparing the response to aircraft impact of a standard reactor building, to that of a reactor building having an independently founded outer shell. In the second scheme, the outer shell is meant to receive the aircraft impact, so that the load will be transmitted to the reactor building internals only by way of the structure-soil-structure system. In both cases, the aircraft impact was postulated to occur on a linear single degree of freedom oscillator which modeled, approximately, the plastification of the impact area. The soil was considered as a half-space with properties corresponding to a medium stiff soil, and modeled by lumped soil springs and dashpots. The reactor internals, inner shell and protective outer shell were modeled with beam elements and concentrated inertias. In modeling the coupled system, soil-structure interaction and structure-to-structure interaction through the soil were represented by a global stiffness matrix corresponding to the three degrees the freedom of each foundation, i.e. horizontal, vertical and rocking. (orig./HP)

  12. Clinch River Breeder Reactor Plant: a building block in nuclear technology

    International Nuclear Information System (INIS)

    McCormack, M.

    1979-01-01

    Interest in breeder reactors dates from the Manhatten Project to the present effort to build the Clinch River Liquid Metal Fast Breeder Reactor (LMFBR) demonstration plant. Seven breeder-type reactors which were built during this time are described and their technological progress assessed. The Clinch River Breeder Reactor Project (CRBRP) has been designed to demonstrate that it can be licensed, can operate on a large power grid, and can provide industry with important experience. As the next logical step in LMFBR development, the project has suffered repeated cancellation efforts with only minor modifications to its schedule. Controversies have developed over the timing of a large-scale demonstration plant, the risks of proliferation, economics, and other problems. Among the innovative developments adopted for the CRBRP is a higher thermal efficiency potential, the type of development which Senator McCormack feels justifies continuing the project. He argues that the nuclear power program can and should be revitalized by continuing the CRBRP

  13. Clinch River Breeder Reactor Plant: a building block in nuclear technology

    Energy Technology Data Exchange (ETDEWEB)

    McCormack, M.

    1979-01-01

    Interest in breeder reactors dates from the Manhatten Project to the present effort to build the Clinch River Liquid Metal Fast Breeder Reactor (LMFBR) demonstration plant. Seven breeder-type reactors which were built during this time are described and their technological progress assessed. The Clinch River Breeder Reactor Project (CRBRP) has been designed to demonstrate that it can be licensed, can operate on a large power grid, and can provide industry with important experience. As the next logical step in LMFBR development, the project has suffered repeated cancellation efforts with only minor modifications to its schedule. Controversies have developed over the timing of a large-scale demonstration plant, the risks of proliferation, economics, and other problems. Among the innovative developments adopted for the CRBRP is a higher thermal efficiency potential, the type of development which Senator McCormack feels justifies continuing the project. He argues that the nuclear power program can and should be revitalized by continuing the CRBRP.

  14. Dismantling and removal of buildings for reactor containment vessel and others

    International Nuclear Information System (INIS)

    Seiki, Yoshihiro

    1996-01-01

    In the practical test of dismantling, after the installed machinery and equipment in respective buildings were removed, the dismantling of the buildings was begun in the state that there was not the contamination of the buildings by removing contaminated concrete, and the buried pipings with the openings being closed remained in the buildings. When contamination did not remain, the control zones of the buildings and their surroundings were canceled, and the dismantling was begun from the upper part of buildings. When contaminated buried pipings remained, the control zones of 1st kind were changed to the control zones of 2nd kind, and the dismantling was begun from the upper part. The dismantling works of buildings were advanced by the procedure of the removal of the pipings penetrating building bodies, the separation and removal of contaminated concrete, the measurement for confirming contamination removal, the change of control zones, the dismantling of buildings, refilling and ground leveling. The removal of buried pipings, the decontamination of buildings, the dismantling and removal of the reactor containment vessel and other buildings, and the results are reported. (K.I.)

  15. Expansion of core melt within the concrete foundation - effects on the reactor building structure

    International Nuclear Information System (INIS)

    Hassmann, K.; Liphardt, W.; Bindseil, P.

    1983-01-01

    The effects of core melt expanding in the concrete fundament of underground-situated nuclear power plants on the structure of the reactor building have been investigated. The radial reduction of the foundation plate is not to exceed 12 m, in case limited additional internal pressure loads of more than 6 bar are to be covered. Greater reductions will endanger the stability of the building. For the assumed load cases, no penetrating cracks will occur. During the subject test, the limiting walls of the building remained leak-proof up to an internal pressure of 2.5 bar. (orig.) [de

  16. Forced vibration test of a nuclear reactor building and its simulation analysis

    International Nuclear Information System (INIS)

    Kobori, T.; Mizuno, N.; Kondo, K.; Niwa, M.; Kobayashi, T.

    1987-01-01

    A forced vibration test was performed on a BWR reactor building of the Hamaoka nuclear power plant unit No. 3 to investigate vibrational characteristics of the building and to verify the adequacy of an analytical model adopted for the aseismic design. The test results show that the fundamental frequency is approximately 3.8 Hz and the damping factor is a large value of approximately 40%. The results of the simulation analysis by means of the Lattice model considering a soil-structure interaction were in good agreement with the test results for both the building and the soil, and consequently the adequacy of the Lattice model was confirmed. (orig.)

  17. Evaluation of the existing reactor building and systems to support TRR renewal program

    International Nuclear Information System (INIS)

    Chia, W.-M.; Chow, T.

    1992-01-01

    Taiwan Research Reactor (TRR), a 40 MW natural uranium heavy-water moderated, light-water cooled NRU type reactor, was one of the research reactors located in the Institute of Nuclear Energy Research (INER). It served as the main test facility to support nuclear R ampersand D activities and radiation services since its first criticality in January 1973. Due to safety and economic reasons, this reactor was shutdown in January 198,8 for a major refurbishment to upgrade its capabilities. After the shutdown of TRR, a large-scale feasibility study aiming to improve the TRR capability was initiated jointly by domestic and foreign experts. Results of this study indicated that, by taking into account various considerations, the innovative approach-one piece reactor removal, i.e. to remove the reactor block by one piece and replaced it with a new one at its original location, as described in a companion paper in this conference, deemed to be the best and most suitable solution to the very complicated problem. This conclusion was true only when the existing reactor building, civil structure as well as major support systems of TRR were judged to be adequate for the new TRR, i.e. TRR-11, use

  18. Effects of different SSI parameters on the floor response spectra of a nuclear reactor building

    International Nuclear Information System (INIS)

    Kabir, A.F.; Bolourchi, S.; Maryak, M.E.

    1991-01-01

    The effects of several critical soil-structure interaction (SSI) parameters on the floor response spectra (FRS) of a typical nuclear reactor building have been examined. These parameters are computation of soil impedance functions using different approaches, scattering effects (reductions in ground motion due to embedment and rigidity of building foundation) and strain dependency of soil dynamic properties. This paper reports that the significant conclusions of the study, which are applicable to a deeply embedded very rigid nuclear reactor building, are as follows: FRS generated without considering scattering effects are highly conservative; differences between FRS, generated considering strain-dependency of soil dynamic properties, and those generated suing low-strain values, are not significant; and the lumped-parameter approach of SSI calculations, which only uses a single value of soil shear modulus in impedance calculations, may not be able to properly compute the soil impedances for a soil deposit with irregularly varying properties with depth

  19. Influence of flexible foundation slab on seismic response of nuclear reactor building

    International Nuclear Information System (INIS)

    Kobatake, M.; Hirashima, S.; Narikawa, M.; Tanaka, H.

    1985-01-01

    In the seismic design of massive and rigid structure such as nuclear reactor building, it is important to properly evaluate the dynamic interaction effect between soil and structure. The authors have carried out the vibration tests of a 1/36th scale model of the BWR type reactor building and its simulation analyses in order to grasp its dynamical behaviour, where the foundation slab was fixed with a shaking table. An analytical study was carried out for a recuced model with flexible foundation slab and semi-infinite soil. Frequency response analyses were then performed by using a FEM model and a lumped-mass model representing soil-structure interaction effect, and studies were made of the influence on building response according to the manner by which the frequency dependency of soil stiffness and the flexibility of foundation slab are treated. (orig.)

  20. Feedback from Westinghouse experience on segmentation of reactor vessel internals - 59013

    International Nuclear Information System (INIS)

    Kreitman, Paul J.; Boucau, Joseph; Segerud, Per; Fallstroem, Stefan

    2012-01-01

    With more than 25 years of experience in the development of reactor vessel internals segmentation and packaging technology, Westinghouse has accumulated significant know-how in the reactor dismantling market. Building on tooling concepts and cutting methodologies developed decades ago for the successful removal of nuclear fuel from the damaged Three Mile Island Unit 2 reactor (TMI-2), Westinghouse has continuously improved its approach to internals segmentation and packaging by incorporating lessons learned and best practices into each successive project. Westinghouse has developed several concepts to dismantle reactor internals based on safe and reliable techniques, including plasma arc cutting (PAC), abrasive water-jet cutting (AWJC), metal disintegration machining (MDM), or mechanical cutting. Westinghouse has applied its technology to all types of reactors covering Pressurized Water Reactors (PWR's), Boiling Water Reactors (BWR's), Gas Cooled Reactors (GCR's) and sodium reactors. The primary challenges of a segmentation and packaging project are to separate the highly activated materials from the less-activated materials and package them into appropriate containers for disposal. Since space is almost always a limiting factor it is therefore important to plan and optimize the available room in the segmentation areas. The choice of the optimum cutting technology is important for a successful project implementation and depends on some specific constraints like disposal costs, project schedule, available areas or safety. Detailed 3-D modeling is the basis for tooling design and provides invaluable support in determining the optimum strategy for component cutting and disposal in waste containers, taking account of the radiological and packaging constraints. Westinghouse has also developed a variety of special handling tools, support fixtures, service bridges, water filtration systems, video-monitoring systems and customized rigging, all of which are required for a

  1. On-line reactor building integrity testing at Gentilly-2 (summary of results 1987-1994)

    International Nuclear Information System (INIS)

    Collins, N.; Lafreniere, P.

    1994-01-01

    In 1987, Hydro-0uebec embarked on an ambitious development program to provide the Gentilly-2 Nuclear Power Station with an effective and practical Reactor Building Containment integrity Test (CIT). In October 1992, the inaugural low pressure (3 kPa(g) nominal) CIT at 100% F.P was performed. The test was conclusive and the CIT was declared In-Service for containment integrity verification on-line. Five subsequent CITs performed in 1993 and 1994 have demonstrated the expected leak rate results and good reliability. The outstanding feature of the CITs is the demonstrated accurary of better than 5% of the measured leak rate. The CIT was developed with the primary goal of demonstrating 'overall' containment availability. Specifically it was designed to detect a 25 mm. diameter leak or hole in the Reactor Building. However, the remarkable CIT accuracy allows reliable detection of a 2 mm. hole. The Gentilly-2 CIT is an innovative approach based on the Temperature Compensation Method (TCM) which uses a reference volume composed of an extensive tubular network of several different diameters. This eliminates the need to track numerous temperature points. A second independent tubular network includes numerous humidity sampling points, thereby enabling the mearurernent of minute pressure variations inside the Reactor Building, independant of the spatial and temporal humidity behaviour. This Gentilly-2 TOM System has been demonstrated to work at both high and low test pressures. The GentiIly-2 design allows the CIT to be performed at a nominal 3 kPa(g) test pressure during a 12-hour period (28 hours total with alignment time) with the reactor at full power. The traditional Reactor Building Pressure Test (RBPT) is typically performed at high pressure (124 kPa(g) in a 5-day critical path window (7 days total with alignment time) during an annual shutdown

  2. Full autonomous monitoring tools inside nuclear reactor building

    International Nuclear Information System (INIS)

    Giraud, A.; Veau, J.

    2009-01-01

    In this paper, we define, design and test a radiation tolerant autonomous monitoring tool for nuclear embedded applications. The goal of the instrumentation system was to record the values of some parameters such as dose, temperature or vibrations appearing inside the containment building of nuclear power plants. The knowledge of these parameters will be a good help for predictive maintenance of the power plant components. For the design of the monitoring tool, we rely on commercial-off-the-shelf (COTS) low power electronic components to use battery-supplied power. A large amount of components starting from discrete transistors or logic units to memories and micro-controllers was associated to define and design a prototype. We then confirm the environment conditions tolerance estimated to up to 2 kGy of total dose and 80 C for temperature by on-line irradiation experiments for individual components and functions and prototypes. Two different sets of about 60 systems were realised as industrial products and then installed in EDF facilities. They were exploited during NPPs operating times of about eighteen months. This large scale experiment was concluded with the inspection of some withdrawn systems. No major degradations were observed showing the efficiency of the hardening method allowing regular use of COTS

  3. Study on vertical seismic response characteristics of deeply embedded reactor building

    International Nuclear Information System (INIS)

    Morishita, H.; Nakamura, N.; Uchiyama, S.; Fukuoka, A.; Ishizaki, M.

    1993-01-01

    This paper describes vertical response characteristics, especially effects of embedment, and analytical methods for seismic design of a deeply embedded reactor building. The influence of embedment on vertical response was found to be minimal by evaluating results of forced vibration tests of a reactor building model and performing simplified analyses. Subsequently, simulation analyses of the forced vibration test and actual earthquake induced response were performed using both the axisymmetric FEM model and the simplified mass and spring model. It was concluded that the analytical models taking the embedment into the consideration closely simulated the observation records, and the omission of embedment in the analyses tended to increase the predicted response which was conservative in respect an actual design consideration. (author)

  4. Response of an offshore nuclear reactor building to simulated tornadic forces

    International Nuclear Information System (INIS)

    Arockiasamy, M.; Reddy, D.V.; Cheema, P.S.

    1977-01-01

    This paper describes the dynamic response to simulated tornadic forces of a nuclear reactor building mounted on a floating platform, moored within a protective breakwater. The time-histories of wind forces are simulated considering the size and layout of the floating nuclear plant (FNP) at several different heights. Due to the strong up-draft and steep pressure gradient at the core of the tornado, the inertial force, usually negligible in winds caused by large scale storms, plays a dominant role in the conversion of wind speed to wind force. Serious underestimation of the wind forces would result if this inertial effect is not considered. The structure considered is a floating platform with the reactor containment shell mounted on it. The eigenvalues and eigenvectors are determined for the platform-building assembly (assumed elastically supported) moored in a basin formed by a rubble mound breakwater. (Auth.)

  5. A simplified method to evaluate restoring force characteristics of reactor buildings

    International Nuclear Information System (INIS)

    Okada, T.; Akino, K.; Ogata, T.

    1987-01-01

    This paper is to propose a discrete beam model analysis to evaluate the non-linear restoring force characteristics of reinforced concrete nuclear reactor buildings macroscopically for the seismic design. In this method, overall deformation of each component such as wall or slab is considered to consist of shearing deformation and bending deformation. A rotational deformation of wall due to the elongation and slip of reinforcing bars from foundation is also considered. Each deformation is evaluated by a discrete beam model based on the non-linear stress-strain relationship estimated by the test results. Examples of verification studies to compare the computed results and the test results on the component scaled models and the scaled reactor building models are also described

  6. A Wireless Monitoring System for Cracks on the Surface of Reactor Containment Buildings

    Directory of Open Access Journals (Sweden)

    Jianguo Zhou

    2016-06-01

    Full Text Available Structural health monitoring with wireless sensor networks has been increasingly popular in recent years because of the convenience. In this paper, a real-time monitoring system for cracks on the surface of reactor containment buildings is presented. Customized wireless sensor networks platforms are designed and implemented with sensors especially for crack monitoring, which include crackmeters and temperature detectors. Software protocols like route discovery, time synchronization and data transfer are developed to satisfy the requirements of the monitoring system and stay simple at the same time. Simulation tests have been made to evaluate the performance of the system before full scale deployment. The real-life deployment of the crack monitoring system is carried out on the surface of reactor containment building in Daya Bay Nuclear Power Station during the in-service pressure test with 30 wireless sensor nodes.

  7. Effects of embedment including slip and separation on seismic SSI response of a nuclear reactor building

    International Nuclear Information System (INIS)

    Saxena, Navjeev; Paul, D.K.

    2012-01-01

    Highlights: ► Both the slip and separation of reactor base reduce with increase in embedment. ► The slip and separation become insignificant beyond 1/4 and 1/2 embedment respectively. ► The stresses in reactor reduce significantly upto 1/4 embedment. ► The stress reduction with embedment is more pronounced in case of tensile stresses. ► The modeling of interface is important beyond 1/8 embedment as stresses are underestimated otherwise. - Abstract: The seismic response of nuclear reactor containment building considering the effects of embedment, slip and separation at soil–structure interface requires modeling of the soil, structure and interface altogether. Slip and separation at the interface causes stress redistribution in the soil and the structure around the interface. The embedment changes the dynamic characteristics of the soil–structure system. Consideration of these aspects allows capturing the realistic response of the structure, which has been a research gap and presented here individually as well as taken together. Finite element analysis has been carried out in time domain to attempt the highly nonlinear problem. The study draws important conclusions useful for design of nuclear reactor containment building.

  8. Data Quality Objective Summary Report for Phase II of the 105-F and DR Reactor Buildings

    International Nuclear Information System (INIS)

    Bauer, R.G.

    1998-01-01

    This data quality objective (DQO) process is to support planning and decision-making activities of Phase II decontamination and decommissioning (D and D) activities for the 105-F and 105-DR Reactor Buildings.The objective of this DQO is to determine the survey and characterization requirements for these rooms to provide the necessary information for worker safety, waste designation, recycle, reuse, and clean landfill disposal decisions during D and D

  9. Precautions against axial fan stall in reactor building to Tianwan NPP

    International Nuclear Information System (INIS)

    Liu Chunlong; Pei Junmin

    2011-01-01

    The paper introduces the mechanism and harm of rotating stall of axial fans, analyzes the necessity for prevention against axial fan stall in reactor building of Tianwan NPP, introduces the precautions, and then makes an assessment on anti-stall effect of flow separators. It can provide reference for model-selection or reconstruction of similar fans in power stations, and for operation and maintenance of axial fans. (authors)

  10. Development of seismic isolation system in vertical direction. Coupled response analysis of reactor building and common deck isolation system

    International Nuclear Information System (INIS)

    Morishita, Masaki

    1997-04-01

    Study was made on the effect of dynamic interaction between reactor building and common deck isolation structure. Dynamic response analyses were made with a coupled and a decoupled model of the reactor building and the isolation structure. The effect of the dynamic interaction was evaluated by comparing these analysis results. The results of the study can be summarized as below; (1) Dynamic coupling effect between reactor building and the common deck isolation structure is not significant. This can be attributed to the fact that the isolation frequency is sufficiently lower than the fundamental frequency of the soil-structure system. (2) The decoupled analysis model was found to give response results with practically sufficient accuracy and slight conservatism. (3) The effect of the response of common deck isolation structure on the response of the reactor building is also negligible. (author)

  11. Response of a NPP reactor building under seismic action with regard to different soil properties

    International Nuclear Information System (INIS)

    Wagenknecht, E.

    1987-01-01

    The object of this investigation is the response of a reactor building on seismic action with systematic variation of the soil stiffness. A thin-walled orthotropic containment shell on varying heavy and rigid foundations is regarded as calculation model. The soil stiffness is simulated by meand of spring elements for horizontal translation and for rocking motions of the building. By the response spectra method the loads of the containment shell are calculated for a horizontal seismic excitation. The investigation is aimed at determining the influence of differentiated soil stiffnesses on the containment action effects and at recognizing the causes for the occuring effects. The results are thoroughly represented by selected quantities of the building's response, the effects from the soil-structure interaction are discussed and the causes of the effects cleary explained. Apossibility is provided for determining critical soil stiffnesses which cause a siginificat intensification effect. The results of the investigations show that both the soil stiffness and structural configuration of the reactor building particulary in case of the substructure being heavy and rigid, exert a decisive on the loading of the superstructure. (orig.)

  12. Release measurements and release from regulatory control of the building containing the Seibersdorf research reactor; Freimessen und Freigabe des Reaktorgebaeudes des Forschungsreaktors Seibersdorf

    Energy Technology Data Exchange (ETDEWEB)

    Steininger, R.; Djuricic, M.; Meyer, F.; Steger, F. [Nuclear Engineering Seibersdorf GmbH (NES), Seibersdorf (Austria)

    2009-07-01

    After 38-year safe operation the 10 MW ASTRA Research Reactor Seibersdorf was switched off in the year 1999. Shortly after the dismantling began and were finalized in the year 2006. The research reactor was housed in a cylindrical building (reactor building, reactor hall) with 26 m diameter and 22.5 m height, which were conceived as reactor containment and which should be free-measured for unlimited applications. Because no neutron activation of the building was expected, the free measurements were confined only of contamination measurements of the building (walls, floors, covers and crane superstructures), which was possibly surface-contaminated by handling for many years with radioactive substances. After free-measurements an application for release of the building was filed at the authorities. After arrival of the release answer the reactor containment was cleared and released from the radiation protection law. The released empty reactor building should be reorganized into a transfer storage of non radioactive material. (orig.)

  13. Safety assessment of A92 reactor building for large commercial aircraft crash

    International Nuclear Information System (INIS)

    Kostov, M.; Henkel, F.O.; Andonov, A.

    2014-01-01

    The current paper presents key elements of the comprehensive analyses of the effects due to a large aircraft collision with the reactor building of Belene NPP in Bulgaria. The reactor building is a VVER A92; it belongs to the third+ generation and includes structural measures for protection against an aircraft impact as standard design. The A92 reactor building implements a double shell concept and is composed of thick RC external walls and an external shell which surrounds an internal pre-stressed containment and the internal walls of the auxiliary building. The malevolent large aircraft impact is considered as a beyond design base accident (Design Extended Conditions, DEC). The main issues under consideration are the structural integrity, the equipment safety due to the induced vibrations, and the fire safety of the entire installation. Many impact scenarios are analyzed varying both impact locations and loading intensity. A large number of non-linear dynamic analyses are used for assessment of the structural response and capacity, including different type of structural models, different finite element codes, and different material laws. The corresponding impact loadings are represented by load time functions calculated according to three different approaches, i.e. loading determined by Riera's method (Riera, 1968), load time function calculated by finite element analysis (Henkel and Klein, 2007), and coupled dynamic analysis with dynamic interaction between target and projectile. Based on the numerical results and engineering assessments the capacity of the A92 reactor building to resist a malevolent impact of a large aircraft is evaluated. Significant efforts are spent on safety assessment of equipment by using an evaluation procedure based on damage indicating parameters. As a result of these analyses several design modifications of structure elements are performed. There are changes of the layout of reinforcement, special arrangements and spatial

  14. Mobile means for the monitoring of atmospheric contamination in a reactor building

    International Nuclear Information System (INIS)

    Marques, S.; Lestang, M.

    2009-01-01

    After having evoked the context and challenges of contamination monitoring when exploiting nuclear reactors, the authors discuss the representativeness of the atmospheric contamination measurement as it depends on the different physicochemical forms of radionuclides present in the circuits. They indicate the different gaseous or aerosol radioactive elements which are monitored within EDF installations. They discuss the incorporation of monitoring means at the installation design level, briefly present the use of beacons inside and outside the reactor building. They describe how monitoring is organized on the basis of alert threshold adjustments: an investigation threshold and an evacuation threshold. They discuss the beacon (or sensor) selection and indicate recommendations for their implementation for optimization purposes. They indicate where these beacons are installed and evoke the experimentation of networked mobile beacons with data remote transmission

  15. Updating of a dynamic finite element model from the Hualien scale model reactor building

    International Nuclear Information System (INIS)

    Billet, L.; Moine, P.; Lebailly, P.

    1996-08-01

    The forces occurring at the soil-structure interface of a building have generally a large influence on the way the building reacts to an earthquake. One can be tempted to characterise these forces more accurately bu updating a model from the structure. However, this procedure requires an updating method suitable for dissipative models, since significant damping can be observed at the soil-structure interface of buildings. Such a method is presented here. It is based on the minimization of a mechanical energy built from the difference between Eigen data calculated bu the model and Eigen data issued from experimental tests on the real structure. An experimental validation of this method is then proposed on a model from the HUALIEN scale-model reactor building. This scale-model, built on the HUALIEN site of TAIWAN, is devoted to the study of soil-structure interaction. The updating concerned the soil impedances, modelled by a layer of springs and viscous dampers attached to the building foundation. A good agreement was found between the Eigen modes and dynamic responses calculated bu the updated model and the corresponding experimental data. (authors). 12 refs., 3 figs., 4 tabs

  16. Determination of floor response spectra for the Brookhaven HFBR reactor building structure

    Energy Technology Data Exchange (ETDEWEB)

    Subudhi, M.; Goradia, H.

    1978-11-01

    In order to perform the dynamic analysis of various structural components of the HFBR reactor building at Brookhaven National Laboratory (BNL) subjected to seismic disturbances, it is necessary to obtain the floor response spectra of the primary structure. The mathematical model includes the four floor levels of the internal structure, the dome, and soil spring effects. The standard time history analysis is adopted to obtain the response spectrum for each floor of the internal structure. This report summarizes the results both in tabular and graphical form for various damping values.

  17. Determination of floor response spectra for the Brookhaven HFBR reactor building structure

    International Nuclear Information System (INIS)

    Subudhi, M.; Goradia, H.

    1978-11-01

    In order to perform the dynamic analysis of various structural components of the HFBR reactor building at Brookhaven National Laboratory (BNL) subjected to seismic disturbances, it is necessary to obtain the floor response spectra of the primary structure. The mathematical model includes the four floor levels of the internal structure, the dome, and soil spring effects. The standard time history analysis is adopted to obtain the response spectrum for each floor of the internal structure. This report summarizes the results both in tabular and graphical form for various damping values

  18. Prediction of prestressing losses for long term operation of nuclear reactor buildings

    Directory of Open Access Journals (Sweden)

    Thillard G.

    2011-04-01

    Full Text Available Prestressed concrete is used in nuclear reactor buildings to guarantee containment and structural integrity in case of an accident. Monitoring and operating experience over 40 years has shown that prestressing losses can be much greater than the design estimation based on the usual standard laws. A method was developed to determine the realistic residual prestress level in structures, in particular for those where no embedded instrumentation was installed, taking into account in situ measurement results rather than design characteristics. The results can enable the owner to justify extending the lifespan while guaranteeing adequate safety and to define and plan adequate maintenance actions.

  19. Study of seismic responses of Candu-3 reactor building using isolator bearings

    International Nuclear Information System (INIS)

    Biswas, J.K.

    1992-01-01

    Seismic isolator bearings are known to increase reliability, reduce cost and increase the potential sitings for nuclear power plants located in regions of high seismicity. High seismic activities in Canada occur mainly in the western coast, the Grand Banks and regions of Quebec along the St. Lawrence river. In Canada, nuclear power plants are located in Ontario, Quebec and New Brunswick where the seismicity levels are low to moderate. Consequently, seismic isolator bearings have not been used in the existing nuclear power plants in Canada. The present paper examines the effect of using seismic isolator bearings in the design for the new CANDU3 which would be suitable for regions having high seismicity. The CANDU3 Nuclear Power Plant is rated at 450 MW of net output power and is a smaller version of its predecessor CANDU6 successfully operating in Canada and abroad. The design of CANDU3 is being developed by AECL CANDU. Advanced technologies for design, construction and plant operation have been utilized. During the conceptual development of the CANDU3 design, various design options including the use of isolator bearings were considered. The present paper presents an overview of seismic isolation technology and summarizes the analytical work for predicting the seismic behavior of the CANDU3 reactor building. A lumped-parameter dynamic model for the reactor building is used for the analysis. The characteristics of the bearings are utilized in the analysis work. The time-history modal analysis has been used to compute the seismic responses. Seismic responses of the reactor building with and without isolator bearings are compared. The isolator bearings are found to reduce the accelerations of the reactor building. As a result, a lower level of seismic qualification for components and systems would be required. The use of these bearings however increases rigid body seismic displacements of the structure requiring special considerations in the layout and interfaces for

  20. SPECIFICATIONS FOR HIGH TEMPERATURE LATTICE TEST REACTOR BUILDING 318 PROJECT CAH-100

    Energy Technology Data Exchange (ETDEWEB)

    Vitro Engineering Company

    1964-07-15

    This is the specifications for the High Temperature Lattice Test Reactor Building 318 and it is divided into the following 21 divisions or chapters: (1) Excavating, Backfill & Grading; (2) Reinforced Concrete; (3) Masonry; (4) Structural Steel & Miscellaneous Metal Items, Contents - Division V; (5) Plumbing, Process & Service Piping; (6) Welding; (7) Insulated Metal Siding; (8) Roof Decks & Roofing; (9) Plaster Partitions & Ceiling; (10) Standard Doors, Windows & Hardware; (11) Shielding Doors; (12) Sprinkler System & Fire Extinguishers, Contents - Division XIII; (13) Heating, Ventilating & Air Conditioning; (14) Painting, Protective Coating & Floor Covering, Contents - Division XV; (15) Electrical; (16) Communications & Alarm Systems; (17) Special Equipment & Furnishings; (18) Overhead Bridge Crane; (19) Prefabricated Steel Building; (20) Paved Drive; and (21) Landscaping & Irrigation Sprinklers.

  1. Incorporating higher order WINKLER springs with 3-D finite element model of a reactor building for seismic SSI analysis

    International Nuclear Information System (INIS)

    Ermutlu, H.E.

    1993-01-01

    In order to fulfill the seismic safety requirements, in the frame of seismic requalification activities for NPP Muehleberg, Switzerland, detailed seismic analysis performed on the Reactor Building and the results are presented previously. The primary objective of the present investigation is to assess the seismic safety of the reinforced concrete structures of reactor building. To achieve this objective requires a rather detailed 3-D finite element modeling for the outer shell structures, the drywell, the reactor pools, the floor decks and finally, the basemat. This already is a complicated task, which enforces need for simplifications in modelling the reactor internals and the foundation soil. Accordingly, all internal parts are modelled by vertical sticks and the Soil Structure Interaction (SSI) effects are represented by sets of transitional and higher order rotational WINKLER springs, i.e. avoiding complicated finite element SSI analysis. As a matter of fact, the availability of the results of recent investigations carried out on the reactor building using diversive finite element SSI analysis methods allow to calibrate the WINKLER springs, ensuring that the overall SSI behaviour of the reactor building is maintained

  2. An approach to build a knowledge base for reactor accident diagnostic expert system

    International Nuclear Information System (INIS)

    Yoshida, K.; Fujii, M.; Fujiki, K.; Yokobayashi, M.; Kohsaka, A.; Aoyagi, T.; Hirota, Y.

    1987-01-01

    In the development of a rule based expert system, one of the key issues is how to acquire knowledge and to build knowledge base (KB). On building the KB of DISKET, which is an expert system for nuclear reactor accident diagnosis developed in JAERI, several problems have been experienced as follows. To write rules is a time consuming task, and it is difficult to keep the objectivity and consistency of rules as the number of rules increase. Further, certainty factors (CFs) must be often determined according to engineering judgment, i.e., empirically or intuitively. A systematic approach was attempted to handle these difficulties and to build an objective KB efficiently. The approach described in this paper is based on the concept that a prototype KB, colloquially speaking an initial guess, should first be generated in a systematic way and then is to be modified and/or improved by human experts for practical use. Statistical methods, principally Factor Analysis, were used as the systematic way to build a prototype KB for the DISKET using a PWR plant simulator data. The source information is a number of data obtained from the simulation of transients, such as the status of components and annunciator etc., and major process parameters like pressures, temperatures and so on

  3. Earthquake response analysis of embedded reactor building considering soil-structure separation and nonlinearity of soil

    International Nuclear Information System (INIS)

    Ichikawa, T.; Hayashi, Y.; Nakai, S.

    1987-01-01

    In the earthquake response analysis for a rigid and massive structure as a nuclear reactor building, it is important to estimate the effect of soil-structure interaction (SSI) appropriately. In case of strong earthquakes, the nonlinearity, such as the wall-ground separation, the base mat uplift of sliding, makes the behavior of the soil-structure system complex. But, if the nuclear reactor building is embedded in a relatively soft ground with surface layer, the wall-ground separation plays the most important role in the response of soil-structure system. Because, it is expected that the base uplift and slide would be less significant due to the effect of the embedment, and the wall-ground friction is usually neglected in design. But, the nonlinearity of ground may have some effect on the wall-ground separation and the response of the structure. These problems have been studied by use of FEM. Others used joint elements between the ground and the structure which does not resist tensile force. Others studied the effect of wall-ground separation with non-tension springs. But the relationship between the ground condition and the effect of the separation has not been clarified yet. To clarify the effect the analyses by FE model and lumped mass model (sway-rocking model) are performed and compared. The key parameter is the ground profile, namely the stiffness of the side soil

  4. Experimental study on joint construction method for aseismatic walls of reactor buildings, (1)

    International Nuclear Information System (INIS)

    Sugita, Kazunao; Mogami, Tatsuo; Ezaki, Tetsuro

    1987-01-01

    On the aseismatic walls of a reactor auxiliary building, many temporary openings are provided at the time of the construction for carrying equipment in later, due to the demand of shortening the construction period. Thus on the aseismatic walls, in most cases there are the joints due to the concrete placed later. As equipment tends to be unitized and become large, the quipment is placed close to the wall having an opening, consequently, the workability is poor, and the standardization of construction method is urgently demanded. The conventional method of closing temporary openings has the problems of safety and connecting reinforcing bars, therefore, the new construction method was proposed. In reactor buildings, the joints of walls are unavoidable, and since those are large scale structures, the joints are numerous. Therefore, at the joint parts, it abandoned and buried frames are used, it is advantageous in the time and cost of joint construction. In both cases, the mechanical properties were confirmed by the fundamental performance test partially modeling the joints and the verifying test modeling the whole walls. In this paper, the test of applying only shearing force to joint models is reported. (Kako, I.)

  5. Incoherent SSI Analysis of Reactor Building using 2007 Hard-Rock Coherency Model

    International Nuclear Information System (INIS)

    Kang, Joo-Hyung; Lee, Sang-Hoon

    2008-01-01

    Many strong earthquake recordings show the response motions at building foundations to be less intense than the corresponding free-field motions. To account for these phenomena, the concept of spatial variation, or wave incoherence was introduced. Several approaches for its application to practical analysis and design as part of soil-structure interaction (SSI) effect have been developed. However, conventional wave incoherency models didn't reflect the characteristics of earthquake data from hard-rock site, and their application to the practical nuclear structures on the hard-rock sites was not justified sufficiently. This paper is focused on the response impact of hard-rock coherency model proposed in 2007 on the incoherent SSI analysis results of nuclear power plant (NPP) structure. A typical reactor building of pressurized water reactor (PWR) type NPP is modeled classified into surface and embedded foundations. The model is also assumed to be located on medium-hard rock and hard-rock sites. The SSI analysis results are obtained and compared in case of coherent and incoherent input motions. The structural responses considering rocking and torsion effects are also investigated

  6. Model tests and numerical analysis on restoring force characteristics of reactor buildings

    International Nuclear Information System (INIS)

    Uchiyama, Y.; Suzuki, S.; Akino, K.

    1987-01-01

    The authors have prepared a nonlinear F.E.M. code called 'SANREF' and made an extensive study for the restoring force characteristics of the inner concrete structures (I/C) of a PWR-type containment vessel and the principal seismic shear walls of a BWR-type reactor building by some series of reduced model tests and simulation analysis for the tests results. The detailed objectives of this study can be summarized as follows: (1) Examine the effectiveness of the configurations of shear walls, reinforcement ratios, shear span ratios (M/Qd) and vertical axial stress by 'partial model test' which simulates some independent shear walls of the PWR-type and BWR-type reactor buildings. (2) Obtain fundamental data of restoring force characteristics of the complex shaped RC structures by 'composite model test' which models are composed of the partial model test specimens. (3) Verify the applicability of analytical methods and constitutive modelings in SANREF code for complex shaped RC structures through nonlinear simulation analysis for the composite model test. (orig./HP)

  7. Building on success. The foreign research reactor spent nuclear fuel acceptance program

    International Nuclear Information System (INIS)

    Huizenga, David G.; Mustin, Tracy P.; Saris, Elizabeth C.; Massey, Charles D.

    1998-01-01

    The second year of implementation of the research reactor spent nuclear fuel acceptance program was marked by significant challenges and achievements. In July 1998, the Department of Energy completed by significant challenges and achievements. In July 1998, the Department of Energy completed its first shipment of spent fuel from Asia via the Concord Naval Weapons Station in California to the Idaho National Engineering and Environmental (INEEL). This shipment, which consisted of three casks of spent nuclear fuel from two research reactors in the Republic of Korea, presented significant technical, legal, and political challenges in the United States and abroad. Lessons learned will be used in the planning and execution of our next significant milestone, a shipment of TRIGA spent fuel from research reactors in Europe to INEEL, scheduled for the summer of 1999. This shipment will include transit across the United States for over 2,000 miles. Other challenges and advances include: clarification of the fee policy to address changes in the economic status of countries during the life of the program; resolution of issues associated with cask certification and the specific types and conditions of spent fuel proposed for transport; revisions to standard contract language in order to more clearly address unique shipping situations; and priorization and scheduling of shipments to most effectively implement the program. As of this meeting, eight shipments, consisting of nearly 2,000 spent fuel assemblies from fifteen countries, have been successfully completed. With the continued cooperation of the international research reactor community, we are committed to building on this success in the remaining years of the program. (author)

  8. Generalized uniform formulae for atmospheric dispersion of activities released from a ventilation stack or from a leaky reactor building

    International Nuclear Information System (INIS)

    Vlachovsky, K.; Veverka, O.; Valenta, V.

    1976-01-01

    Relations are given for the calculation of the atmospheric transport and dispersion of industrial gaseous wastes released from the stacks of factories, power plants and nuclear power plants. Modified formulae are derived for stack disposal from a small stack, also applicable in calculating the gaseous waste release from a leaky reactor. Uniform generalized formulae are presented serving the calculation of both high and short stack disposals as well as of reactor building leakages. (L.O.)

  9. Experimental and analytical studies of a deeply embedded reactor building model considering soil-building interaction. Pt. 3

    International Nuclear Information System (INIS)

    Tanaka, H.

    1983-01-01

    The paper describes the dynamic charachteristics of a deeply embedded reactor building model obtained from the forced vibration tests, earthquake observations and simulation analysis. The earthquake records of the structure and the surrounding soil were examined by using soil-building interaction model as used in the analyses of the forced vibration tests. It is considered that the response of the structure will be influenced by the seismic behaviour of the soil layer as the seismic wave is input to the bedrock of the soil-structure interaction model in the earthquake response analysis. Therefore, dynamic properties of the soil layer during earthquakes were investigated in detail, and applied to the seismic simulation analysis using soil-structure interaction model. Many earthquake records have been obtained since June, 1976 when the earthquake observation system was first established. From these, eight of them which had comparatively large acceleration values were used to investigate the transfer properties of soil layer. Besides, transfer functions computed using in-situ measurement shearing wave velocity showed good agreement with those of the earthquake records. The records of the Miyagiken-oki earthquake of February 20, 1978 (magnitude 6.7) was selected as an example for performing simulation analysis. The simulation analysis are as follows: (1) In the seismic simulation analysis using soil-structure interaction modal, computed results will be in good agreement with the observed ones, when the transfer function of soil layer is properly estimated. (2) Judging from the transfer function of soil layer with the characteristics that the modal damping value decreases gradually at a higher modal frequency, it is found that ddamping of soil-layer can be simulated more adequately by introducing external damping system together with structural damping. (orig./HP)

  10. Assessment of extent and degree of thermal damage to polymeric materials in the Three Mile Island Unit 2 reactor building

    International Nuclear Information System (INIS)

    Alvares, N.J.

    1986-01-01

    This paper describes assumptions and procedures used to perform thermal damage analysis caused by post loss-of-coolant-accident (LOCA) hydrogen deflagration at Three Mile Island Unit 2 Reactor. Examination of available photographic evidence yields data on the extent and range of thermal and burn damage. Thermal damage to susceptible material in accessible regions of the reactor building was distributed in non-uniform patterns. No clear explanation for non-uniformity was found in examined evidence, e.g., burned materials were adjacent to materials that appear similar but were not burned. Because these items were in proximity to vertical openings that extend the height of the reactor building, the authors assume the unburned materials preferentially absorbed water vapor during periods of high, local steam concentration. A control pendant from the polar crane located in the top of the reactor building sustained asymmetric burn damage of decreasing degree from top to bottom. Evidence suggests the polar-crane pendant side that experienced heaviest damage was exposed to intense radiant energy from a transient fire plume in the reactor containment volume. Simple hydrogen-fire-exposure tests and heat transfer calculations approximate the degree of damage found on inspected materials from the containment building and support for an estimated 8% pre-fire hydrogen

  11. Assessment of extent and degree of thermal damage to polymeric materials in the Three Mile Island Unit 2 Reactor building

    International Nuclear Information System (INIS)

    Alvares, N.J.

    1985-06-01

    This paper describes assumptions and procedures used to perform thermal damage analysis caused by post loss-of-coolant-accident (LOCA) hydrogen deflagration at Three Mile Island Unit 2 Reactor. Examination of available photographic evidence yields data on the extent and range of thermal and burn damage. Thermal damage to susceptible material in accessible regions of the reactor building was distributed in non-uniform patterns. No clear explanation for non-uniformity was found in examined evidence, e.g., burned materials were adjacent to materials that appear similar but were not burned. Because these items were in proximity to vertical openings that extend the height of the reactor building, we assume the unburned materials preferentially absorbed water vapor during periods of high, local steam concentration. A control pendant from the polar crane located in the top of the reactor building sustained asymmetric burn damage of decreasing degree from top to bottom. Evidence suggests the polar-crane pendant side that experienced heaviest damage was exposed to intense radiant energy from a transient fire plume in the reactor containment volume. Simple hydrogen-fire-exposure tests and heat transfer calculations approximate the degree of damage found on inspected materials from the containment building and support for an estimated 8% pre-fire hydrogen

  12. Damage of reactor buildings occurred at the Fukushima Daiichi accident. Focusing on sequence leading to hydrogen explosions

    International Nuclear Information System (INIS)

    Naito, Masanori

    2011-01-01

    Fukushima Daiichi accident discharged enormous radioactive materials confined inside into the environment due to hydrogen explosions occurred at reactor buildings and forced many people to live the refugee life. This article described overview of Great East Japan Earthquake, specifications of Fukushima Daiichi nuclear power plants, sequence of plant status after earthquake occurrence and computerized simulation of plant behavior of Unit 1 leading to core melt and hydrogen explosion. Simulation results with estimated and assumed conditions showed water level decreased to bottom of reactor core after 4 hrs and 15 minutes passed, core melt started after 6 hrs and 49 minutes passed, failure of core support plate after 7 hrs and 18 minutes passed and through failure of penetration at bottom of pressure vessel after 7 hrs and 25 minutes passed. Hydrogen concentration at operating floor of reactor building of Unit 1 would be 15% accumulated and the pressure would amount to about 5 bars after hydrogen explosion if reactor building did not rupture with leak-tight structure. Since reactor building was not pressure-proof structure, walls of operating floor would rupture before 5 bars attained. (T. Tanaka)

  13. An approach to build a knowledge base for reactor diagnostic system using statistical method

    International Nuclear Information System (INIS)

    Yokobayashi, Masao; Matsumoto, Kiyoshi; Kohsaka, Atsuo

    1988-01-01

    In the development of a rule-based expert system, one of the key issues is how to acquire knowledge and to build a knowledge base. When the knowledge base of DISKET was built, which is an expert system for nuclear reactor accident diagnosis developed in Japan Atomic Energy Research Institute, several problems have been experienced. To write rules is a time-consuming task, and it was difficult to keep the objectivity and consistency of rules as the number of rules increased. Certainty factors must be determined often according to engineering judgement, i.e. empirically or intuitively. A systematic approach was attempted to cope with these difficulties and to build efficiently an objective knowledge base. The approach described in this paper is based on the concept that a prototype knowledge base, colloquially speaking an initial guess, should first be generated in a systematic way, then it is modified or improved by human experts for practical use. Factor analysis was used as the systematic way. DISKET system, the procedure of building a knowledge base, and the verification of the approach are reported. (Kako, I.)

  14. On detonation dynamics in hydrogen-air-steam mixtures: Theory and application to Olkiluoto reactor building

    International Nuclear Information System (INIS)

    Silde, A.; Lindholm, I.

    2000-02-01

    This report consists of the literature study of detonation dynamics in hydrogen-air-steam mixtures, and the assessment of shock pressure loads in Olkiluoto 1 and 2 reactor building under detonation conditions using the computer program DETO developed during this work at VTT. The program uses a simple 1-D approach based on the strong explosion theory, and accounts for the effects of both the primary or incident shock and the first (oblique or normal) reflected shock from a wall structure. The code results are also assessed against a Balloon experiment performed at Germany, and the classical Chapman-Jouguet detonation theory. The whole work was carried out as a part of Nordic SOS-2.3 project, dealing with severe accident analysis. The initial conditions and gas distribution of the detonation calculations are based on previous severe accident analyses by MELCOR and FLUENT codes. According to DETO calculations, the maximum peak pressure in a structure of Olkiluoto reactor building room B60-80 after normal shock reflection was about 38.7 MPa if a total of 3.15 kg hydrogen was assumed to burned in a distance of 2.0 m from the wall structure. The corresponding pressure impulse was about 9.4 kPa-s. The results were sensitive to the distance used. Comparison of the results to classical C-J theory and the Balloon experiments suggested that DETO code represented a conservative estimation for the first pressure spike under the shock reflection from a wall in Olkiluoto reactor building. Complicated 3-D phenomena of shock wave reflections and focusing, nor the propagation of combustion front behind the shock wave under detonation conditions are not modeled in the DETO code. More detailed 3-D analyses with a specific detonation code are, therefore, recommended. In spite of the code simplifications, DETO was found to be a beneficial tool for simple first-order assessments of the structure pressure loads under the first reflection of detonation shock waves. The work on assessment

  15. EVALUATION OF THE DURABILITY OF THE STRUCTURAL CONCRETE OF REACTOR BUILDINGS AT SRS

    Energy Technology Data Exchange (ETDEWEB)

    Duncan, A.; Reigel, M.

    2011-02-28

    The Department of Energy (DOE) intends to close 100-150 facilities in the DOE complex using an in situ decommissioning (ISD) strategy that calls for grouting the below-grade interior volume of the structure and leaving the above-grade interior open or demolishing it and disposing of it in the slit trenches in E Area. These closures are expected to persist and remain stable for centuries, but there are neither facility-specific monitoring approaches nor studies on the rate of deterioration of the materials used in the original construction or on the ISD components added during closure (caps, sloped roofs, etc). This report will focus on the evaluation of the actual aging/degradation of the materials of construction used in the ISD structures at Savannah River Site (SRS) above grade, specifically P & R reactor buildings. Concrete blocks (six 2 to 5 ton blocks) removed from the outer wall of the P Reactor Building were turned over to SRNL as the first source for concrete cores. Larger cores were received as a result of grouting activities in P and R reactor facilities. The cores were sectioned and evaluated using microscopy, x-ray diffraction (XRD), ion chromatography (IC) and thermal analysis. Scanning electron microscopy shows that the aggregate and cement phases present in the concrete are consistent with the mix design and no degradation mechanisms are evident at the aggregate-cement interfaces. Samples of the cores were digested and analyzed for chloride ingress as well as sulfate attack. The concentrations of chloride and sulfate ions did not exceed the limits of the mix design and there is no indication of any degradation due to these mechanisms. Thermal analysis on samples taken along the longitudinal axis of the cores show that there is a 1 inch carbonation layer (i.e., no portlandite) present in the interior wall of the reactor building and a negligible carbonation layer in the exterior wall. A mixed layer of carbonate and portlandite extends deeper into the

  16. Nonlinear seismic response analysis of embedded reactor buildings based on the substructure approach in time domain

    International Nuclear Information System (INIS)

    Hasegawa, M.; Nakai, S.; Watanabe, T.

    1985-01-01

    A practical method for elasto-plastic seismic response analysis is described under considerations of nonlinear material law of a structure and dynamic soil-structure interaction. The method is essentially based on the substructure approach of time domain analysis. Verification of the present method is carried out for typical BWR-MARK II type reactor building which is embedded in a soil, and the results are compared with those of the frequency response analysis which gives good accuracy for linear system. As a result, the present method exhibits sufficient accuracy. Furthermore, elasto-plastic analyses considering the soil-structure interaction are made as an application of the present method, and nonlinear behaviors of the structure and embedment effects are discussed. (orig.)

  17. FEM analysis of foundation raft for 500 MWe pressurized heavy water reactor building

    International Nuclear Information System (INIS)

    Kulkarni, N.N.; Goray, J.S.; Joshi, M.H.; Paramasivam, V.

    1989-01-01

    Foundation raft supports the containment structure and internals for 500 MWe PHW reactor building. It also serves as bottom envelop of the containment structure. In view of this, the design of foundation raft assumes great importance. The foundation raft is subjected to various load, most significant of them are dead load of structure, equipment loads transferred through a system of floors, walls and structural steel columns, pressure load during accident conditions, seismic loads, earth pressure, uplift due to buoyancy loads, foundation reaction etc. In order to achieve optimum design, the detailed structural analysis is required to be performed methodically and in most realistic manner. Finite element methods which have come in vogue with the developments in digital computers can be successfully applied in this area. The paper describes the above methods in detail for the analysis of foundation raft for the various load combinations required to be considered for safe and optimum design

  18. Investigation of cable deterioration in the containment building of the Savannah River Nuclear Reactor

    International Nuclear Information System (INIS)

    Gillen, K.T.; Clough, R.L.; Jones, L.H.

    1982-08-01

    This report describes an investigation of the deterioration of polyethylene and polyvinylchloride cable materials which occurred in the containment building of the Savannah River nuclear reactor located at Aiken, South Carolina. Radiation dosimetry and temperature mapping data of the containment area indicated that the maximum dose experienced by the cable materials was only 2.5 Mrad at an average operating temperature of 43 0 C. Considering this relatively moderate environment, the amount of material degradation seemed surprising. To understand these findings, an experimental program was performed on the commercial polyethylene and polyvinylchloride materials used at the plant to investigate their degradation behavior under combined γ-radiation and elevated temperature conditions. It is established that the material deterioration at the plant resulted from radiation-induced oxidation and that the degradation rate can be correlated with local levels of radiation intensity in the containment area

  19. Pre-service proof pressure and leak rate tests for the Qinshan CANDU project reactor buildings

    International Nuclear Information System (INIS)

    Petrunik, K.J.; Khan, A.; Ricciuti, R.; Ivanov, A.; Chen, S.

    2003-01-01

    The Qinshan CANDU Project Reactor Buildings (Units 1 and 2) have been successfully tested for the Pre-Service Proof Pressure and Integrated Leak Rate Tests. The Unit 1 tests took place from May 3 to May 9, 2002 and from May 22 to May 25, 2002, and the Unit 2 tests took place from January 21 to January 27, 2003. This paper discusses the significant steps taken at minimum cost on the Qinshan CANDU Project, which has resulted in a) very good leak rate (0.21%) for Unit 1 and excellent leak rate (0.130%) for Unit 2; b) continuous monitoring of the structural behaviour during the Proof Pressure Test, thus eliminating any repeat of the structural test due to lack of data; and c) significant schedule reduction achieved for these tests in Unit 2. (author)

  20. Forced vibration tests on the reactor building of a nuclear power station, 1

    International Nuclear Information System (INIS)

    Takeda, Toshikazu; Tsunoda, Tomohiko; Wakamatsu, Kunio; Kaneko, Masataka; Nakamura, Mitsuru; Kunoh, Toshio; Murahashi, Hisahiro

    1988-01-01

    Tsuruga Unit No.2 Nuclear Power Station of the Japan Atomic Power Company is the first PWR-type 4-loop plant constructed in Japan with a prestressed concrete containment vessel (PCCV). This report describes forced vibration tests carried out on the reactor building of this plant. The following were obtained as results: (1) The results of the forced vibration tests corresponded well on the whole with design values. (2) The vibration characteristics of the PCCV observed in the tests after prestressing are no different from the ones before prestressing. This shows that the vibration properties of the PCCV are practically independent of prestressing loads. (3) A seismic response analysis of the design basis earthquake was made on the design model reflecting the test results. The seismic safety of the plant was confirmed by this analysis. (author)

  1. Decommissioning of the ASTRA research reactor - dismantling the auxiliary systems and clearance and reuse of the buildings

    International Nuclear Information System (INIS)

    Meyer, F.; Steger, F.; Steininger, R.

    2008-01-01

    The paper presents work performed in the last phase of the decommissioning of the ASTRA research reactor at the Austrian Research Centers Seibersdorf. Dismantling the pump room installations and the ventilation system, as well as the clearance of the buildings is described. Some conclusions and summary data regarding the time table, material management, and the cost of the entire project are also presented. (author)

  2. Answers to questions about removing krypton from the Three Mile Island, Unit 2 reactor building. Public information report

    International Nuclear Information System (INIS)

    1980-05-01

    This document presents answers to frequently asked questions about the probable effects of controlled releases of the krypton presently contained within the reactor building of Three Mile Island, Unit 2. Also answered are questions about alternative means for removing the krypton

  3. The effect of vertical earthquake component on the uplift of the nuclear reactor building

    International Nuclear Information System (INIS)

    Kobayashi, Toshio

    1986-01-01

    During a strong earthquake, the base mat of a nuclear reactor building may be lifted partially by the response overturning moment. And it causes geometrical nonlinear interaction between the base mat and rock foundation beneath it. In order to avoid this uplift phenomena, the base mat and/or plan of the building is enlarged in some cases. These special design need more cost and/or time in construction. In the evaluation of the uplift phenomena, a parameter ''η'' named ''contact ratio'' is used defined as the ratio of compression stress zone area of base mat for total area of base mat. Usually this contact ratio is calculated under the combination of the maximum overturning moment obtained by the linear earthquake response analysis and the normal force by the gravity considering the effect of the vertical earthquake component. In this report, the effect of vertical earthquake component for the uplift phenomena is studied and it concludes that the vertical earthquake component gives little influence on the contact ratio. In order to obtain more reasonable contact retio, the nonlinear rocking analysis subjected to horizontal and vertical earthquake motions simultaneously is proposed in this report. As the second best method, the combination of the maximum overturning moment obtained by linear analysis and the normal force by only the gravity without the vertical earthquake effect is proposed. (author)

  4. Removal plan for Shippingport pressurized water reactor core 2 blanket fuel assemblies form T plant to the canister storage building

    Energy Technology Data Exchange (ETDEWEB)

    Lata

    1996-09-26

    This document presents the current strategy and path forward for removal of the Shippingport Pressurized Water Reactor Core 2 blanket fuel assemblies from their existing storage configuration (wet storage within the T Plant canyon) and transport to the Canister Storage Building (designed and managed by the Spent Nuclear Fuel. Division). The removal plan identifies all processes, equipment, facility interfaces, and documentation (safety, permitting, procedures, etc.) required to facilitate the PWR Core 2 assembly removal (from T Plant), transport (to the Canister storage Building), and storage to the Canister Storage Building. The plan also provides schedules, associated milestones, and cost estimates for all handling activities.

  5. Neutron dosimetry in the Three-Mile Island Unit 2 reactor cavity with solid-state track recorders

    International Nuclear Information System (INIS)

    Gold, R.; Roberts, J.H.; Ruddy, F.H.; Preston, C.C.; McElroy, W.N.; Rao, S.V.; Greenborg, J.; Fricke, V.R.

    1986-01-01

    Solid-state track recorder (SSTR) neutron dosimetry has been conducted in the Three-Mile Island Unit 2 (TMI-2) reactor cavity, for nondestructive assessment of the fuel distribution. Two axial stringers were deployed in the annular gap with 17 SSTR dosimeters located on each stringer. SSTR experimental results reveal that neutron streaming, upward from the bottom of the reactor cavity region, dominates the observed neutron intensity. These absolute thermal neutron flux observations are consistent with the presence of a significant amount of fuel debris lying at the bottom of the reactor vessel. A conservative lower bound estimated from these SSTR data implies that at least 2 tonnes of fuel, which is roughly 4 fuel assemblies, is lying at the bottom of the vessel. This existence of significant neutron streaming also explains the high count rate observed with the source range monitors that are located in the TMI-2 reactor cavity. (author)

  6. Reactor

    International Nuclear Information System (INIS)

    Toyama, Masahiro; Kasai, Shigeo.

    1978-01-01

    Purpose: To provide a lmfbr type reactor wherein effusion of coolants through a loop contact portion is reduced even when fuel assemblies float up, and misloading of reactor core constituting elements is prevented thereby improving the reactor safety. Constitution: The reactor core constituents are secured in the reactor by utilizing the differential pressure between the high-pressure cooling chamber and low-pressure cooling chamber. A resistance port is formed at the upper part of a connecting pipe, and which is connect the low-pressure cooling chamber and the lower surface of the reactor core constituent. This resistance part is formed such that the internal sectional area of the connecting pipe is made larger stepwise toward the upper part, and the cylinder is formed larger so that it profiles the inner surface of the connecting pipe. (Aizawa, K.)

  7. Assessment of thermal damage to polymeric materials by hydrogen deflagration in the Three Mile Island Unit 2 Reactor Building

    International Nuclear Information System (INIS)

    Alvares, N.J.

    1985-05-01

    Thermal damage to susceptible material in accessible regions of the reactor building was distributed in non-uniform patterns. No clear explanation for non-uniformity was found in examined evidence, e.g., burned materials were adjacent to materials that appear similar but were not burned. Because these items were in proximity to vertical openings that extend the height of the reactor building, we assume the unburned materials preferentially absorbed water vapor during periods of high, local steam concentration. Simple hydrogen-fire-exposure tests and heat transfer calculations duplicate the degree of damage found on inspected materials from the containment building. These data support estimated 8% pre-fire hydrogen concentration predictions based on various hydrogen production mechanisms

  8. Transport of breeder reactor-fire-generated sodium oxide aerosols for building-wake-dominated meteorology

    Energy Technology Data Exchange (ETDEWEB)

    Fields, D.E.; Cooper, A.C.; Miller, C.W.

    1987-02-01

    This report describes the methodology used and results obtained in efforts to estimate the sodium aerosol concentrations at air intake ports of a liquid-metal cooled, fast-breeder nuclear reactor. An earlier version of this methodology has been previously discussed (Fields and Miller, 1985). A range of wind speeds from 2 to 10 m/s is assumed, and an effort is made to include building wake effects which, in many cases, dominate the dispersal of aerosols near buildings. For relatively small release rates, on the order of 1 to 10 kg/s, the plume rise is small and estimates of aerosol concentrations are derived using the methodology of Wilson and Britter (1982), which describes releases from surface vents. For release rates on the order of 100 kg/s much higher release velocities are expected, and plume rise is considered. An effective increase in release height is computed using the Split-H methodology with a parameterization suggested by Ramsdell (1983), and the release source strength is transformed to rooftop level. Evaluation of the acute release aerosol concentration is then based on the methodology for releases from a surface release of this transformed source strength. For a horizontal release, a methodology is developed to chart the plume path as a function of release and site meteorology parameters. Results described herein must be regarded as maximum aerosol concentrations, based on models derived from generic wind tunnel studies. More accurate and site-specific results may be obtained through wind tunnel simulations and through simulating emissions from release points other than those assumed here.

  9. Linked Data Reactor: a Framework for Building Reactive Linked Data Applications

    NARCIS (Netherlands)

    Khalili, Ali

    2016-01-01

    This paper presents Linked Data Reactor (LD-Reactor or LD-R) as a framework for developing exible and reusable User Interface components for Linked Data applications. LD-Reactor utilizes Facebook's ReactJS components, Flux architecture and Yahoo's Fluxible framework for isomorphic Web applications.

  10. Build-up and decay of fuel actinides in the fuel cycle of nuclear reactors

    International Nuclear Information System (INIS)

    Tasaka, Kanji; Kikuchi, Yasuyuki; Shindo, Ryuichi; Yoshida, Hiroyuki; Yasukawa, Shigeru

    1976-05-01

    For boiling water reactors, pressurized light-water reactors, pressure-tube-type heavy water reactors, high-temperature gas-cooled reactors, and sodium-cooled fast breeder reactors, uranium fueled and mixed-oxide fueled, each of 1000 MWe, the following have been studied: (1) quantities of plutonium and other fuel actinides built up in the reactor, (2) cooling behaviors of activities of plutonium and other fuel actinides in the spent fuels, and (3) activities of plutonium and other fuel actinides in the high-level reprocessing wastes as a function of storage time. The neutron cross section and decay data of respective actinide nuclides are presented, with their evaluations. For effective utilization of the uranium resources and easy reprocessing and high-level waste management, a thermal reactor must be fueled with uranium; the plutonium produced in a thermal reactor should be used in a fast reactor; and the plutonium produced in the blanket of a fast reactor is more appropriate for a fast reactor than that from a thermal reactor. (auth.)

  11. Integrated leak rate testing of the fast flux test facility reactor containment building

    International Nuclear Information System (INIS)

    James, E.B.; Farabee, O.A.; Bliss, R.J.

    1978-01-01

    The initial Integrated Leak Rate Test (ILRT) of the Fast Flux Test Facility containment building was performed from May 27 to June 2, 1978. The test was conducted in air with systems vented and with the containment recirculating coolers in operation. 10 psig and 5 psig tests were run using the absolute pressure test method. The measured leakage rates were .033% Vol/24 hr. and -.0015% Vol/24 hrs. respectively. Subsequent verification tests at both 10 psig and 5 psig proved that the test equipment was operating properly and it was sensitive enough to detect leaks at low pressures. This ILRT was performed at a lower pressure than any previous ILRT on a reactor containment structure in the United States. While the initial design requirements for ice condenser containments called for a part pressure test at 6 psig, the tests were waived due to the apparent statistical problems of data analysis and the repeatability of the data itself at such low pressure. In contrast to this belief, both the 5 and 10 psig ILRT's were performed in a successful manner at FFTF

  12. Characterization of an aerosol sample from the auxiliary building of the Three Mile Island reactor.

    Science.gov (United States)

    Kanapilly, G M; Stanley, J A; Newton, G J; Wong, B A; DeNee, P B

    1983-11-01

    Analyses for radioisotopic composition and dissolution characteristics were performed on an aerosol filter sample collected for a week by an air sampler located in the auxiliary building of the Three Mile Island nuclear reactor. The major radioisotopes found on the filter were 89Sr, 90Sr, 134Cs and 137Cs. Greater than 90% of both 89-90Sr and 134-137Cs dissolved within 48 hr in an in vitro test system. Scanning electron microscopic analyses showed the presence of respirable size particles as well as larger particles ranging up to 10 micron in diameter. The major matrix components were Fe, Ca, S, Mg, Al and Si. Although the radionuclides were present in a heterogeneous matrix, they were in a soluble form. This information enables a better evaluation of bioassay data and predictions of dose distribution resulting from an inhalation exposure to this aerosol. Further, the combination of techniques used in this study may be applicable to the characterization of other aerosols of unknown composition.

  13. Experimental and numerical determination of the dynamic properties of the reactor building of Atucha II NPP

    International Nuclear Information System (INIS)

    Ceballos, M.A.; Car, E.J.; Prato, T.A.; Prato, C.A.; Alvarez, L.M.; Godoy, A.R.

    1995-01-01

    Determination of the dynamic properties of the reactor building of Atucha II NPP is carried out in order to: i) Obtain valuable information for seismic qualification of the plant, and ii) Assess some procedures for testing and analysis that are used in the process of seismic evaluation of existing nuclear facilities founded on Quaternary soil deposits. Both steady state and impulsive dynamic tests were performed but attention is centered here in tile techniques used to determine natural frequencies and modal damping ratios with impulsive tests. Numerical analyses were performed by means of a 3-D model model of the superstructure together with foundation stiffness coefficients derived in a separate paper from steady state vibration tests, and also from analysis with a 2-D F.E. model of the soil layers capable of approximating the 3-D features of the problem. The computed foundation stiffness coefficients are compared both with those obtained from the tests and from an axisymmetric F.E. model; results indicate that foundation stiffness coefficients calculated with F.E. models with soil parameters given by laboratory tests performed on cored samples are significantly lower than those given by the steady state vibration tests. (author)

  14. Re-evaluation of floor response spectra of reactor building for Daya Bay NPP

    International Nuclear Information System (INIS)

    Li Zhongcheng; Li Zhongxian

    2006-01-01

    The seismic analysis of nuclear island of Daya Bay Nuclear Power Plant (NPP) was just in accordance with the approaches in RCC-G standard for the model M310 in France, in which the simplified impedance matrix method was employed for the consideration of soil's function. In this paper the more sophisticated 3D half-space continuum impedance method based on the Green functions is used to analyze the function of soil. In addition, multi-group of input time histories was used in the seismic response analysis in the existing design and their average of responses for each group was taken as the design basis. The same multi-group of input time histories was used in the seismic response analysis in this study, but the average and enveloped value of responses for each case are calculated respectively to account for the uncertainty of input motions. Focused on the above two issues, the seismic responses of the reactor building are calculated and the floor response spectra (FRS), a very important data for the design of nuclear system and equipment, are generated and then compared with the corresponding design values, so that the evaluation and discuss on the design method and results are conducted. Some useful conclusions hopefully to provide some important references to the assessment of seismic safety margin for the NPP in operation are taken. (authors)

  15. Pressure test at the reactor building of the Embalse Nuclear Power Plant (CNE)

    International Nuclear Information System (INIS)

    Coutsiers, E.E.; Perrino, J.; Moreno, C.; Batistic, J.A.; Lolis, R.R.; Aviles, A.

    1991-01-01

    Upon request by the Licensing Authority, the reactor building (RB) in a nuclear power plant must be submitted to pressure tests. One of these tests is to be performed before startup and, then, a test must be carried out every 5 years in operation. The pre-operational tests took place in August 1981, under two values of relative pressure: 1.266 kg/cm 2 and 0.422 kg/cm 2 . Operational tests must only be made at the lower pressure and their objective is to verify that the loss speed remains within the range indicated in the corresponding technical specification. The first operational test was performed in August 1989. The personnel of the CNE took care of the preparation of the Work Plan, of aligning the various systems contained in the RB, of pressurization, of monitoring localized tightedness, of depressurization and of the general and quality control of the test. The measurements were carried out by the CISME (Center of Metrology Research and Service) of the National Institute of Industrial Technology (INTI) , which did also supply the necesary instruments and the data collection system. There is also a description of the work performed before the test, of the calculation method used for assessing the loss rate, of the test sequencies and of the results obtained. (Author) [es

  16. Limit the effects of secondary circuit water or steam piping breaks in the reactor building

    International Nuclear Information System (INIS)

    Nachev, N.

    2001-01-01

    The existing design of the WWER-1000 Model 320 does not include provisions against the local mechanical effects of pipe ruptures of the secondary system piping. This situation may lead to accidental effects beyond the design basis of the plant in case of a postulated secondary pipe rupture event. The aim of the present safety enhancement measure is to overcome this safety deficit, that means to carry out some analyses and to suggest protection measures, by which the specified design basis of the plant concerning secondary circuit design basis accidents will be assured. The systems to be considered include the main steam lines (MSL) and the main feedwater lines (MFWL) in the safety related system areas. These areas are the system portions, which are located in the reactor building (containment and room A820 outside the containment). The pipe rupture effects to be considered include the local effects, that means pipe whip impact and jet forces on the adjacent equipment and structures, as well as reaction forces due to blowdown thrust forces and pressure waves in the broken piping system. (author)

  17. The 3D-FEM modeling of the LAES unit 1 reactor building for extreme external effects

    International Nuclear Information System (INIS)

    1999-01-01

    In order to study the extreme external effects, three dimensional model was applied to study the effects of aircraft crash and gas explosion on the reactor building of Leningrad-1 NPP which is modelled by finite element method. The crash loads taken into account were from Cessna civil airplane crash with impact velocity of 360 km/h and maximum impact force of 7 MN and the Phantom military airplane crash with impact velocity of 215 km/h and maximum impact force of 110 MN. The gas explosion load was assumed to affect the reactor building from one side parallel to one of the global coordinate axes of the model. The conclusion drawn from the obtained results is as follows: the intersections stiffen the structure considerably. In lower part, where many intersections exist, displacements were significantly smaller. Thus, the lower parts can resist the investigated loads such as high speed military aircraft crash loads much better than the upper part

  18. Experimental study on 1/30 scale model of reinforced concrete reactor building under cyclic lateral loading

    International Nuclear Information System (INIS)

    Fukada, Yasuo; Hirashima, Shinichi; Shohara, Ryoichi

    1980-01-01

    At the time of carrying out the aseismatic design of BWR type reactor buildings, in particular the design to the limit earthquake for design (S2 earthquake), it is necessary to grasp the load-deformation relationship of the buildings not only in elastic region but also extending to plastic region. The BWR type reactor buildings are the complex structures composed of the thick walls of box type and cylindrical-conical type. As the first stage of the experimental approach for grasping various properties of such structures to earthquake force, three kinds of the main elements of the buildings resisting earthquakes were taken out individually, and three 1/30 scale models were subjected to horizontal loading experiment. In this paper, the concept of this experiment, tested models, experimental method and a part of the experimental results are reported. The experiment took about 14 months from the design of the tested models to the completion of experiment. At present, the data analysis is continued. The stress intensity corresponding to the limit earthquake for design was lower than 1/2.5 as compared with the ultimate end, and sufficient margin remained. The deformation at that time was sufficiently small, and the number and the width of cracks were also small. The buildings seemed to accomplish the functions sufficiently. (Kako, I.)

  19. Confirmatory Survey Results for the Reactor Building Dome Upper Structural Surfaces, Rancho Saco Nuclear Generating Station, Herald, California

    International Nuclear Information System (INIS)

    Wade C. Adams

    2006-01-01

    Results from a confirmatory survey of the upper structural surfaces of the Reactor Building Dome at the Rancho Seco Nuclear Generating Station (RSNGS) performed by the Oak Ridge Institute for Science and Education for the NRC. Also includes results of interlaboratory comparison analyses on several archived soil samples that would be provided by RSNGS personnel. The confirmatory surveys were performed on June 7 and 8, 2006

  20. Decommissioning of the ASTRA research reactor: Dismantling the auxiliary systems and clearance and reuse of the buildings

    Directory of Open Access Journals (Sweden)

    Meyer Franz

    2008-01-01

    Full Text Available The paper presents work performed in the last phase of the decommissioning of the ASTRA research reactor at the Austrian Research Centers Seibersdorf. Dismantling the pump room installations and the ventilation system, as well as the clearance of the buildings is described. Some conclusions and summary data regarding the timetable, material management, and the cost of the entire project are also presented.

  1. Effects of uncertainties in soil-structure interaction models on nonlinear seismic responses of nuclear reactor buildings

    International Nuclear Information System (INIS)

    Mizuno, J.; Namba, H.; Komura, T.; Tuchiya, Y.; Saitoh, M.

    1993-01-01

    Effects of uncertainties in materials and modeling on seismic responses of nuclear reactor buildings are relatively small if responses remain in the elastic range of structures. However, under extremely severe earthquakes, responses are expected to go into nonlinear ranges, and responses may exhibit wider scatters due to the uncertainties and variabilities in materials and nonlinear characteristics of shear walls. Thus, it may be quite important to evaluate the effects of the uncertainties on nonlinear responses, especially, the maximum shear strain response, which is one of the most critical responses to the seismic safety of reinforced concrete shear walls and to reactor buildings as a whole. To this end, the sensitivities of nonlinear responses, mainly shear strain responses, to the uncertainties and variabilities of materials and nonlinear characteristics of shear walls arc evaluated for a reactor building of BWR Mark I type; then, the variabilities of nonlinear responses due to the uncertainties and variabilities are evaluated on the basis of the first-order approximation of response statistics as well as numerical simulations

  2. Dynamic containment of gaseous effluents in the auxiliary buildings and reinjection of liquid effluents from these buildings back into the reactor building for 900 MWe PWRs under accident condition

    International Nuclear Information System (INIS)

    Demoulin, F.; Collinet, J.; Nguyen, C.

    1987-04-01

    Examination of the lessons to be learned from the accident of the Three Mile Island nuclear power plant on 20 March 1979 led the French Safety Authorities and EDF (Electricite de France) to adopt a series of measures intended to improve the performance of the containment of French PWRs, especially in the event of accident. Among the measures adopted, two of them contribute to the upgrading of the containment of nuclear island buildings, by reducing radioactivity constraints inside these buildings and by limiting radioactive releases into the environment. These are: (1) dynamic containment of auxiliary buildings likely to be contaminated following an accident, (2) reinjection back into the reactor building of liquid effluents arising in the auxiliary buildings. In this paper we shall discuss, for each measure, the approach to the problem and describe the arrangements made to arrive at a satisfactory solution [fr

  3. Advances in Multi-Sensor Scanning and Visualization of Complex Plants: the Utmost Case of a Reactor Building

    Science.gov (United States)

    Hullo, J.-F.; Thibault, G.; Boucheny, C.

    2015-02-01

    In a context of increased maintenance operations and workers generational renewal, a nuclear owner and operator like Electricité de France (EDF) is interested in the scaling up of tools and methods of "as-built virtual reality" for larger buildings and wider audiences. However, acquisition and sharing of as-built data on a large scale (large and complex multi-floored buildings) challenge current scientific and technical capacities. In this paper, we first present a state of the art of scanning tools and methods for industrial plants with very complex architecture. Then, we introduce the inner characteristics of the multi-sensor scanning and visualization of the interior of the most complex building of a power plant: a nuclear reactor building. We introduce several developments that made possible a first complete survey of such a large building, from acquisition, processing and fusion of multiple data sources (3D laser scans, total-station survey, RGB panoramic, 2D floor plans, 3D CAD as-built models). In addition, we present the concepts of a smart application developed for the painless exploration of the whole dataset. The goal of this application is to help professionals, unfamiliar with the manipulation of such datasets, to take into account spatial constraints induced by the building complexity while preparing maintenance operations. Finally, we discuss the main feedbacks of this large experiment, the remaining issues for the generalization of such large scale surveys and the future technical and scientific challenges in the field of industrial "virtual reality".

  4. Geological and geotechnical aspects of the foundation pit of Kaiga atomic power plant reactor building 2, Kaiga, Uttara Kannada district, Karnataka

    International Nuclear Information System (INIS)

    Katti, Vinod J.; Shah, V.L.; Pande, A.K.

    2014-01-01

    In India Nuclear Power Plants are constructed as per the guidelines laid by IAEA and AERB. Before concrete is poured into reactor building pits, they are systematically mapped and Iithostructural maps are prepared for pit base and side walls. The constraints noticed are carefully attended with geotechnical solutions and remedies to make foundation safe for the entire period of reactor life. Similarly, pit of Kaiga Reactor Building II was systematically mapped for circular base and side walls. Geo-engineering solutions like scrapping out loose, foliated schistose patches, scooping out soft altered zones, filling with grouting, rock-bolting rock segments with major joints and fractures for stopping seepage points were suggested. (author)

  5. Building on knowledge base of sodium cooled fast breeder reactors to develop materials technology for fusion reactors

    International Nuclear Information System (INIS)

    Raj, B.; Rao, B.S.

    2007-01-01

    Full text of publication follows: There are strong synergies between fission and fusion research. Similarities extend to candidate structural materials, methods of theoretical and experimental investigations, high operating temperatures, fast neutron exposure and similar underlying physical phenomena. Enhancing the cross-fertilization of fission and fusion R and D programmes would mutually benefit the development of both kinds of nuclear reactors for sustainable nuclear energy. Materials used in fusion energy systems must be fully code qualified for high-temperature applications that include creep, low cycle fatigue, thermo-mechanical fatigue and creep-fatigue interaction. Evaluation of various mechanical properties of reduced activation ferritic-martensitic alloys are underway for establishing reliable data bases for development of a suitable code and further optimization and redesign of the alloys for fusion reactor components. The alloys 316L (N) and derived variants of Mod.9Cr1Mo steel are anticipated for use in fusion systems and the RCC-MR code would serve as a backbone for the initial design of various components. In this respect, the efforts made in India for developing these materials, their characterization and applicability of RCC-MR code are highlighted. The salient features of consortium approach and the experience that has led to manufacturing of main and safety vessel components from 316L (N) with close dimensional tolerances, thick section forged components and seamless tubes of Modified 9Cr-1Mo steel for steam generators will be presented. The finite element analysis has been used for optimal design of die and punch for the near net shape forming of petals of main vessel of PFBR. The safe temperature and strain rate domains were established using dynamic materials model for forming of 316L (N) and 9Cr1Mo steels by various industrial processes. Welds are weak links in the design of both fission and fusion systems. Detailed investigations have led to

  6. Dynamic analysis of the reactor building for soft (Kozloduy) and hard (Temelin) soil conditions and different seismic loading

    International Nuclear Information System (INIS)

    Krutzik, N.

    1995-01-01

    Analyses were conducted for the reactor building to determine the dynamic responses of the coupled system, soil and structure and the forces in the characteristic structural members. This report summarizes the results of structural dynamic analyses derived for soft and hard soil conditions by the modal time history method using substructure models as well as (for soft soil conditions) in the frequency domain using complex (coupled) models of the soil and the structure. The mathematical model of the reactor building is represented as a lumped mass beam model. The capabilities of the soil were represented by means of global frequency independent springs and dampers (substructure models) or by an appropriate final element model. The results of the above-mentioned analysis presented in this report comprise in particular the maximum values of accelerations, displacements and internal forces as well as the acceleration response spectra for the relevant building regions. The time domain (modal time history) calculations were performed for real soil conditions which corresponds to the site Kozloduy (soft) and Temelin (hard). As seismic input data the corresponding free-field data here been used. The dynamic response obtained for the soft-soil conditions using both type of (substructure and complex) models were compared and demonstrated in one plot. Similar comparison were performed for the results obtained for soft and hard soil conditions

  7. A SEISMIC DESIGN OF NUCLEAR REACTOR BUILDING STRUCTURES APPLYING SEISMIC ISOLATION SYSTEM IN A HIGH SEISMICITY REGION –A FEASIBILITY CASE STUDY IN JAPAN-

    Directory of Open Access Journals (Sweden)

    TETSUO KUBO

    2014-10-01

    Full Text Available A feasibility study on the seismic design of nuclear reactor buildings with application of a seismic isolation system is introduced. After the Hyogo-ken Nanbu earthquake in Japan of 1995, seismic isolation technologies have been widely employed for commercial buildings. Having become a mature technology, seismic isolation systems can be applied to NPP facilities in areas of high seismicity. Two reactor buildings are discussed, representing the PWR and BWR buildings in Japan, and the application of seismic isolation systems is discussed. The isolation system employing rubber bearings with a lead plug positioned (LRB is examined. Through a series of seismic response analyses using the so-named standard design earthquake motions covering the design basis earthquake motions obtained for NPP sites in Japan, the responses of the seismic isolated reactor buildings are evaluated. It is revealed that for the building structures examined herein: (1 the responses of both isolated buildings and isolating LRBs fulfill the specified design criteria; (2 the responses obtained for the isolating LRBs first reach the ultimate condition when intensity of motion is 2.0 to 2.5 times as large as that of the design-basis; and (3 the responses of isolated reactor building fall below the range of the prescribed criteria.

  8. The intelligent customer: considerations around build-own-operate business and licensing models for small modular reactors in Canada

    International Nuclear Information System (INIS)

    Jones, K.

    2014-01-01

    An organization planning a proposal for a build-own-operate business model needs to address expanded licensee responsibilities under this model, associated regulatory impacts and how this affects their role as an 'intelligent customer'. This is particularly important for cases where builder-owner-operators plan to manufacture factory-fuelled designs and ship them to a site for installation and operation. The primary responsibility for safe conduct of licensed activities rests with the licensee. A build-own-operate model expands the scope of licensed activities to include design, manufacturing, transport, construction, and operation. The licensee must be able to demonstrate they are qualified to conduct all licensed activities including sufficient competent resources within the licensee's organization to oversee('Intelligent Customer') any work it commissions externally and the subsequent flow down through of the supply chain. This paper examines aspects that organizations need to assess the suitability of approaches that it may take to maintain in-house expertise for the control and oversight of licensed activities at all times. It considers the approach to identification of: core capabilities the licensee would need to understand its safety case under a build-own-operate model to manage licensed activities in accordance with requirements under the Nuclear Safety and Control Acta licensee's 'intelligent customer' capabilities in particular around understanding, specifying, overseeing and accepting work undertaken on its behalf by contractors. While this paper is focused on small modular reactors, being an intelligent customer applies to large commercial or research reactors equally; the size of reactor is immaterial.

  9. The intelligent customer: considerations around build-own-operate business and licensing models for small modular reactors in Canada

    Energy Technology Data Exchange (ETDEWEB)

    Jones, K., E-mail: kenneth.jones@cnsc-ccsn.gc.ca [Canadian Nuclear Safety Commission, Ottawa, Ontario (Canada)

    2014-07-01

    An organization planning a proposal for a build-own-operate business model needs to address expanded licensee responsibilities under this model, associated regulatory impacts and how this affects their role as an 'intelligent customer'. This is particularly important for cases where builder-owner-operators plan to manufacture factory-fuelled designs and ship them to a site for installation and operation. The primary responsibility for safe conduct of licensed activities rests with the licensee. A build-own-operate model expands the scope of licensed activities to include design, manufacturing, transport, construction, and operation. The licensee must be able to demonstrate they are qualified to conduct all licensed activities including sufficient competent resources within the licensee's organization to oversee('Intelligent Customer') any work it commissions externally and the subsequent flow down through of the supply chain. This paper examines aspects that organizations need to assess the suitability of approaches that it may take to maintain in-house expertise for the control and oversight of licensed activities at all times. It considers the approach to identification of: core capabilities the licensee would need to understand its safety case under a build-own-operate model to manage licensed activities in accordance with requirements under the Nuclear Safety and Control Acta licensee's 'intelligent customer' capabilities in particular around understanding, specifying, overseeing and accepting work undertaken on its behalf by contractors. While this paper is focused on small modular reactors, being an intelligent customer applies to large commercial or research reactors equally; the size of reactor is immaterial.

  10. Seismic response analysis of nuclear reactor buildings under consideration of soil-structure interaction with torsional behavior

    International Nuclear Information System (INIS)

    Mizuno, N.; Iida, T.; Tsushima, Y.; Araki, T.; Nojima, O.

    1977-01-01

    In this paper, the seismic response analysis is described in detail for estimating the soil-structure interaction effects with the torsional behavior. The analytical method is firstly shown for estimating the stiffness of reactor building by the bending-shear and torsion theory of the thin wall sections in regard to the behavior of structure. The three-dimensional behavior of structure can be obtained more briefly and simply by the proposed method. Secondly, the dynamical soil-foundation coefficient for estimating the dissipation of vibrational energy on the ground is derived by H. Tajimi's theory which is based on a solution of the propagation of seismic waves caused by point excitation on the surface of the elastic half-space medium. The above results give the vibrational impedances of the soil-foundation corresponding to the static soil coefficient, which is defined to the excitation force in the frequency domain. In order to analyze to the equivalues of reactor building, the authors thirdly attempt to approximate the dynamic soil-foundation coefficient as the frequency transfer function of displacement. The complex damping is used for more suitably estimating the elastic structural damping effects of structure. The regression analysis of many degrees of freedom is fourthly attempted for estimating the natural periods annd equivalent viscous damping ratios directly from the experimental results by the forced vibrational test performed in 1974. The analytical results are finally shown for simulating and comparing with the above-mentioned experimental results

  11. Reactor

    International Nuclear Information System (INIS)

    Fujibayashi, Toru.

    1976-01-01

    Object: To provide a boiling water reactor which can enhance a quake resisting strength and flatten power distribution. Structure: At least more than four fuel bundles, in which a plurality of fuel rods are arranged in lattice fashion which upper and lower portions are supported by tie-plates, are bundled and then covered by a square channel box. The control rod is movably arranged within a space formed by adjoining channel boxes. A spacer of trapezoidal section is disposed in the central portion on the side of the channel box over substantially full length in height direction, and a neutron instrumented tube is disposed in the central portion inside the channel box. Thus, where a horizontal load is exerted due to earthquake or the like, the spacers come into contact with each other to support the channel box and prevent it from abnormal vibrations. (Furukawa, Y.)

  12. Reactor

    International Nuclear Information System (INIS)

    Evans, R.M.

    1976-01-01

    Disclosed is 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. 1 claim, 16 figures

  13. Causes and effects of coolant parameters oscillations on NPP with light water reactor

    International Nuclear Information System (INIS)

    Proskuriakov, K.N.

    2004-01-01

    Paper describes some original academic and engineering R and D results carried out in NPPD MPEI (TU) which that time advanced the current requirements of nuclear power. Acoustic method of early detection of steam cavity appearance in WWER and PWR reactor core had been published before the accident at TMI-2. Specialists recognized this method capability for prevention of one of the most severe accident. Subject of our researches is clearing up of the nature and frequencies of coolant parameters oscillations in operational and emergency situations. Calculated eigenfrequencies of self-excited coolant pressure oscillations based on carry out models had been verified by measurements on NPP with WWER-1000. Spectrum identification (most possible in presence) is provided. The dependence of coolant pressure oscillation amplitudes from reactor power level was revealed. As the probable mechanism of dynamic stresses occurrence in the hot collector's welded line SG-4, the parametrical resonance concept is offered. Bring forward CSGDPS that in particular shows the strong influence of hydraulic resistances on two phases flow stability. As CSGDPS much more than in real NPP systems (due to difference in hydraulic resistances) it is reasonable to realize of PSIS experimental units' reconstruction. The objective of our initiative is to build the computer code for calculating the coolant pressure eigenfrequencies oscillations at normal and emergency operation. This code could provide possibility of NPP vibroacoustical certification. Also the objective of R and D is theoretically substantiation and determination the control actions for resonance prevention between coolant and structures (fuel assemblies, reactor internals and equipment), the service life prolongation, the improvement of NPP reliability and the safety. (author)

  14. Remote maintenance in the building of the reactor of power plants

    International Nuclear Information System (INIS)

    Bonin, R.

    1984-01-01

    Examples describing the different operations requiring remote control for reactor maintenance are given. These operations include: refueling machines (for closure stud, vessel flange cleaning, screwing plug for channel head, swimming pool decontamination) in-service inspection machines (MIS, spider for eddy current testing of steam generator, television) and routine or accidental maintenance (leak detection in water boxes, maintenance spider, opening or closing primary manways, decontamination manipulators and various automatic control devices) [fr

  15. Experimental and analytical studies for a BWR nuclear reactor building. Evaluation of soil-structure interaction behaviour

    International Nuclear Information System (INIS)

    Mizuno, N.; Tsushima, Y.

    1975-01-01

    This paper evaluates the spatial characteristics of dynamic properties, especially soil-structure interaction behaviour, of the BWR nuclear building by experimental and analytical studies. It is well known that the damping effects in soil-structure interaction are remarkable on the building with short periods by the dissipation of vibrational energy to the soil. The authors have previously reported an analytical method for estimating the damping effects the properties of which are characterized as follows: 1) The complex damping is used, because the so-called structural damping may be more suitable for estimating the damping effects of an elastic structure. 2) H. Tajimi's theory is used for estimating the dynamical soil-foundation stiffness with the dissipation of vibrational energy on the elastic half-space soil. In this paper, an approximate explanation is presented in regard to the more developmental mathematical method for estimating the damping effects than the above-mentioned previous method, which is 'Modes Superposition Method for Multi-Degrees of Freedom System' with the constant complex stiffness showing the structural damping effects and the dynamical soil-foundation stiffness approximated by the linear or quadratic functions of the eigenvalues. An approximate explanation is presented in regard to the experimental results of the No. 1 reactor building (BWR) of Hamaoka Nuclear Power Station, The Chubu Electric Power Co., Ltd. (Auth.)

  16. Building

    Directory of Open Access Journals (Sweden)

    Ashwani kumar

    2015-08-01

    This paper discusses existing development scenario and issues to accommodate future development in hill towns located in Indian Himalayan region, also highlights the state of existing building regulations through an in-depth study of building regulations in major hill towns, and briefly discuses possible approaches to change existing building regulations for achieving contextually appropriate development.

  17. 3-dimensional earthquake response analysis of embedded reactor building using hybrid model of boundary elements and finite elements

    International Nuclear Information System (INIS)

    Muto, K.; Motosaka, M.; Kamata, M.; Masuda, K.; Urao, K.; Mameda, T.

    1985-01-01

    In order to investigate the 3-dimensional earthquake response characteristics of an embedded structure with consideration for soil-structure interaction, the authors have developed an analytical method using 3-dimensional hybrid model of boundary elements (BEM) and finite elements (FEM) and have conducted a dynamic analysis of an actual nuclear reactor building. This paper describes a comparative study between two different embedment depths in soil as elastic half-space. As the results, it was found that the earthquake response intensity decreases with the increase of the embedment depth and that this method was confirmed to be effective for investigating the 3-D response characteristics of embedded structures such as deflection pattern of each floor level, floor response spectra in high frequency range. (orig.)

  18. Project management lessons learned from building the Wendelstein 7-x stellerator fusion research reactor

    Energy Technology Data Exchange (ETDEWEB)

    Freire-Gormaly, M.; Gittens, A.; Zhang, L., E-mail: m.freire.gormaly@utoronto.ca, E-mail: antonio.gittens@mail.utoronto.ca, E-mail: lavender.zhang@outlook.com [University of Toronto, Toronto, ON (Canada)

    2015-07-01

    Wendelstein 7-X (W7-X) is the world's largest 'stellerator' nuclear fusion reactor being commissioned in Greifswald, Germany. It will inform the international fusion energy test device (ITER). The complexity of W7- X added challenges since industrial expertise to manufacture components did not exist. The construction was completed eight years behind schedule and almost 100% over budget. Key take-away lessons in project management were revealed from W7-X which can be applied to any nuclear project. These lessons are aligned with the project management knowledge areas of schedule, stakeholder, procurement, scope, schedule, cost, communication, risk, quality, human resources and procurement management. (author)

  19. ORNL fusion power demonstration study: arguments for a vacuum building in which to enclose a fusion reactor

    International Nuclear Information System (INIS)

    Werner, R.W.

    1976-12-01

    Fusion reactors as presently contemplated are excessively complicated, are virtually inaccessible for some repairs, and are subject to frequent loss of function. This dilemma arises in large part because the closed surface that separates the ''hard'' vacuum of the plasma zone from atmospheric pressure is located at the first wall or between blanket and shield. This closed surface is one containing hundreds to thousands of linear meters of welds or mechanical seals which are subject to radiation damage and cyclic fatigue. In situ repair is extremely difficult. This paper examines the arguments favoring the enclosing of the entire reactor in a vacuum building and thus changing the character of this closed surface from one requiring absolute vacuum integrity to one of high pumping impedance. Two differentially pumped vacuum zones are imagined, one clean zone for the plasma and one for the balance of the volume. Both would be at substantially the same pressure. Other advantages for the vacuum enclosure are also cited and discussed

  20. Earthquake response analyses of boiling water type reactor buildings to the vertical ground motions

    International Nuclear Information System (INIS)

    Minami, T.; Ohkawa, I.; Tohdo, M.; Iida, S.; Taga, A.; Kojima, M.

    1985-01-01

    Vertical response analyses of the BWR buildings have been carried out for some typical analytical models to confirm their validities in evaluating maximum responses of the structure. Elastic-plastic analyses of the roof structure are discussed next, under some consideration of gravitational effects in the vertical earthquake responses. Finally, the interacting effects between horizontal and vertical ground excitations on the base-mat up-lifts are studied by a certain conventional method. (orig.)

  1. Round table: The New Constructions; New Build; IAEA Services Nuclear Power Reactor Technologies Construction and Operation

    International Nuclear Information System (INIS)

    Choho, Tarik; Ward, Tony; Koshy, Thomas; )

    2014-01-01

    The first round-table of Atoms for the Future 2014 focused on the topic of 'The new construction'. Attendees had the opportunity to see Tarik Choho, Chief Commercial Executive Officer at AREVA, Tony Ward, Partner and Head of Power and Utilities at EY, as well as Thomas Koshy, Head of Nuclear Power technology Development Section at the IAEA, share their respective point of view and discuss together. They first started with their own presentation before sitting together at the discussion panel. Tarik Choho explained that there are many opportunities for new builds around the world, and especially in new comer countries. Thomas Koshy brought his expert point of view as an international regulator. He showed that the technology landscape also includes new designs as well as non-electricity generation uses. Those two aspects could bring both opportunities and challenges for licensing. Tony Ward showed how difficult it was to finance such a capitalistically intensive project as a new build, and how one could try to find an electricity retail price which is acceptable for both the general private consumer and the plant operator. All speakers agreed that preparation was core to a nuclear new build project. They also pointed out that every project involves more and more actors, and that it could be challenging to let everyone have the appropriate say

  2. Gas and water permeability of concrete for reactor buildings--prototype scale specimens

    International Nuclear Information System (INIS)

    Mills, R.H.

    1987-02-01

    The permeability testing was performed on four concrete cylinders, 0.25 m in diameter and 2 m long, modelling the wall-thickness of reactor containment structures on the prototype scale. Tests were performed on the cylinders before and after artificial induction of longitudinal cracks, intented to model defects developing after some period of adverse service conditions. Permeability increased greatly with the introduction of longitudinal cracks in the concrete, and was also affected by moisture content and casting direction. The influence of reinforcing steel could not be resolved within the bounds of experimental variability. Ultrasound measurements were taken on each cylinder before and after cracking, and a correlation between increased permeability and lowered Ultrasonic Pulse Velocity was observed. Ultrasonic Pulse Velocity measurements thus show promise as a means of continuous monitoring of the integrity of the concrete barrier in service

  3. Earthquake response analysis of embedded reactor building considering soil-structure separation and nonlinearity of soil

    International Nuclear Information System (INIS)

    Ichikawa, T.; Hayashi, Y.; Nakai, S.

    1987-01-01

    The effect of the wall-ground separation depends on the relation between the fundamental frequency of the SSI system and that of the surface layer. The maximum accelerations of the upper floors are increased if the side soil is soft. The building shear force is decreased below the ground level if the fundamental frequency of the SSI system is nearly equal to that of the surface layer. The floor response spectra are slightly increased in the high frequency range. Yielding of the soil occurred only in case that the side soil is soft, and the yield zone was restricted in the upper part of the surface layer. Therefore, the material nonlinearity did not affect the results so much. The results of the sway-rocking model (lumped mass model) analysis showed good agreements with those of the FEM models. (orig./HP)

  4. Underwater plasma arc cutting in Three Mile Island's reactor

    International Nuclear Information System (INIS)

    McGough, M.S.; Knetl, G.J.; Austin, W.E.

    1989-01-01

    On March 28, 1979, the Pennsylvania Three Mile Island nuclear power plant Unit 2 (TMI-2) suffered a partial fuel-melt accident. During this accident, over 20,000 lb of molten fuel flowed through holes melted through the baffle plates and through the lower-core support assembly (LCSA). The molten fuel subsequently resolidified in the bottom of the reactor vessel. The lower-core support assembly of the TMI-2 reactor was not structurally damaged during the accident. In order to permit defueling of that region of the core, the LCSA was cut to permit access. A five-axis teleoperator was developed to deliver plasma arc cutting, rotary grinding and abrasive waterjet cutting of end effectors to the LCSA. Complex geometry sectioning was completed in a mock-up facility at chemistry and pressure conditions simulating those of the vessel, prior to actual in-vessel operations. In-vessel activities began in early May 1988 and were completed on April 11, 1989. This paper presents the details of the in-vessel cutting efforts

  5. Reactor building 3D-model for evaluating the pressures on concrete regularization and foundation waterproofing membrane

    International Nuclear Information System (INIS)

    Mello Junior, Glauco J.T.; Cardoso, Tarcisio de F.; Prates, Carlos L.M.

    2009-01-01

    Angra dos Reis site in Brazil has already 2 operating PWR NPPs. Unit 3, with identical design to Unit 2, also a 1350 MW PWR, is expected to have its construction started in 2009. This new plant shall be founded directly on sound rock. The first step is to prepare this rock surface with a concrete regularization and a foundation waterproofing membrane. This study presents a 3D model approach of the corresponding reactor building to verify the maximum pressure acting on this surface. The 3D model permits to show a more realistic pressure distribution at every foundation specific detail. A static analysis is performed using ANSYS Mechanical Release 11.1. Dead weight, permanent and live loads, Safe Shutdown Earthquake (SSE) combined with Burst Pressure Wave (BPW) from the Feedwater Tank (SSB=SSE+BPW) and differences of temperature are taken into account. Considering all foundation nodes , the pressure distribution on the waterproofing membrane for each load case is obtained for vertical and horizontal directions, which corresponds to compression and tangential reaction loads. The maximum values occur in distinct positions for each load case. The maximum results are obtained according to DIN 25449 (2008) load combination criteria. The results are compared to a simplified analysis performed before, showing a good agreement in global values. (author)

  6. Radiation impact caused by the rupture of a radioactive tank within the Reactor Auxiliary Building of Angra 2

    International Nuclear Information System (INIS)

    Passos, Erivaldo Mario dos; Alves, Antonio Sergio de Martin

    2002-01-01

    This paper aims to show the methodology, the parameters and some results of the radionuclide migration simulation in order to determine the radiation impact to the biosphere due to an accidental radionuclide release associated with the rupture of a radioactive tank within the Reactor Auxiliary Building of Angra 2. After tank rupture, the radionuclides are supposed to reach the sea via the aquifer of the Angra 2 site. This radiological impact is evaluated with the aid of the activity concentration at the sea and dose received by members of the public. Activity concentration for each radionuclide is calculated according to the ANSI/ANS - 2.17 - 1980, which shows the methodology for calculation of activity concentration in the aquifer in case of accidental radionuclide releases of nuclear power plants, whereas the dose calculation follows recognized international procedures. The migration analysis for the mentioned radionuclides is performed through the aquifer and allows to estimate the maximum activity concentration near the sea boundary and the annual dose to the member of the public. Based on the safety analysis performed for the investigated case one can conclude the annual dose impact is lower than that corresponding to one year of normal operation of the Angra 2 plant. (author)

  7. Design response spectra-compliant real and synthetic GMS for seismic analysis of seismically isolated nuclear reactor containment building

    Energy Technology Data Exchange (ETDEWEB)

    Ali, Ahmer [ENVICO Consultants Co. Ltd., Seoul (Korea, Republic of); Abu-Hayah, Nadin; Kim, Doo Kie [Civil and Environmental Engineering, Kunsan National University, Gunsan (Korea, Republic of); Cho, Sung Gook [Innose Tech Co., Ltd., Incheon (Korea, Republic of)

    2017-06-15

    Due to the severe impacts of recent earthquakes, the use of seismic isolation is paramount for the safety of nuclear structures. The diversity observed in seismic events demands ongoing research to analyze the devastating attributes involved, and hence to enhance the sustainability of base-isolated nuclear power plants. This study reports the seismic performance of a seismically-isolated nuclear reactor containment building (NRCB) under strong short-period ground motions (SPGMs) and long-period ground motions (LPGMs). The United States Nuclear Regulatory Commission-based design response spectrum for the seismic design of nuclear power plants is stipulated as the reference spectrum for ground motion selection. Within the period range(s) of interest, the spectral matching of selected records with the target spectrum is ensured using the spectral-compatibility approach. NRC-compliant SPGMs and LPGMs from the mega-thrust Tohoku earthquake are used to obtain the structural response of the base-isolated NRCB. To account for the lack of earthquakes in low-to-moderate seismicity zones and the gap in the artificial synthesis of long-period records, wavelet-decomposition based autoregressive moving average modeling for artificial generation of real ground motions is performed. Based on analysis results from real and simulated SPGMs versus LPGMs, the performance of NRCBs is discussed with suggestions for future research and seismic provisions.

  8. Proposed and existing passive and inherent safety-related structures, systems, and components (building blocks) for advanced light-water reactors

    Energy Technology Data Exchange (ETDEWEB)

    Forsberg, C.W.; Moses, D.L.; Lewis, E.B.; Gibson, R.; Pearson, R.; Reich, W.J.; Murphy, G.A.; Staunton, R.H.; Kohn, W.E.

    1989-10-01

    A nuclear power plant is composed of many structures, systems, and components (SSCs). Examples include emergency core cooling systems, feedwater systems, and electrical systems. The design of a reactor consists of combining various SSCs (building blocks) into an integrated plant design. A new reactor design is the result of combining old SSCs in new ways or use of new SSCs. This report identifies, describes, and characterizes SSCs with passive and inherent features that can be used to assure safety in light-water reactors. Existing, proposed, and speculative technologies are described. The following approaches were used to identify the technologies: world technical literature searches, world patent searches, and discussions with universities, national laboratories and industrial vendors. 214 refs., 105 figs., 26 tabs.

  9. Proposed and existing passive and inherent safety-related structures, systems, and components (building blocks) for advanced light-water reactors

    International Nuclear Information System (INIS)

    Forsberg, C.W.; Moses, D.L.; Lewis, E.B.; Gibson, R.; Pearson, R.; Reich, W.J.; Murphy, G.A.; Staunton, R.H.; Kohn, W.E.

    1989-10-01

    A nuclear power plant is composed of many structures, systems, and components (SSCs). Examples include emergency core cooling systems, feedwater systems, and electrical systems. The design of a reactor consists of combining various SSCs (building blocks) into an integrated plant design. A new reactor design is the result of combining old SSCs in new ways or use of new SSCs. This report identifies, describes, and characterizes SSCs with passive and inherent features that can be used to assure safety in light-water reactors. Existing, proposed, and speculative technologies are described. The following approaches were used to identify the technologies: world technical literature searches, world patent searches, and discussions with universities, national laboratories and industrial vendors. 214 refs., 105 figs., 26 tabs

  10. Sampling and Analysis Plan for Disposition of the Standing Legacy Wastes in the 105-B, -D, -H, -KE, and -KW Reactor Buildings

    International Nuclear Information System (INIS)

    McGuire, J.J.

    1999-01-01

    This sampling and analysis plan (SAP) presents the rationale and strategy for the sampling and analysis activities that support disposition of legacy waste in the Hanford Site's 105-B, 105-D, 105-H, 105-KE, 105-KW Reactor buildings. For the purpose of this SAP, legacy waste is identified as any item present in a facility that is not permanently attached to the facility and is easily removed without the aid of equipment larger than a standard forklift

  11. Sampling and Analysis Plan for Disposition of the Standing Legacy Wastes in the 105-B, -D, -H, -KE, and -KW Reactor Buildings

    International Nuclear Information System (INIS)

    McGuire, J. J.

    1999-01-01

    This sampling and analysis plan (SAP) presents the rationale and strategy for the sampling and analysis activities that support disposition of legacy waste in the Hanford Site's 105-B, 105-D, 105-H,105-KE, 105-KW Reactor buildings. For the purpose of this SAP, legacy waste is identified as any item present in a facility that is not permanently attached to the facility and is easily removed without the aid of equipment larger than a standard forklift

  12. Analysis of the impact of an aircraft crash on underground concrete ducts with protective slab at reactor buildings

    International Nuclear Information System (INIS)

    Kotulla, B.; Hansson, V.

    1977-01-01

    In this paper different types of idealization for a dynamic analysis of underground concrete ducts with protective slab are discussed and compared. Ducts between reactor and control building of a nuclear power plant are to be designed for loadings produced by an aircraft crash. These ducts have a height of about three to four meters and are two to eight meters wide. They are designed with a protective slab about 1.5 m in thickness at ground level and with an intermediate layer of earth of about one meter in thickness. An analysis has to take into account the combined effects of a protective slab with a relatively thin intermediate layer of earth and the underlaying duct and layer of soil with the nonlinear behavior of concrete due to cracking. For describing this behavior two types of idealization were made. One type is a continuum type calculation which describes the slab, the soil and the duct by finite elements. In the other type of idealization a model consisting of springs and lumped masses is used. The protective slab and the intermediate layer of earth may be described as a plate on elastic foundation. The behavior of the cracked part of the plate and the part of earth layer beneath and loads transferred to the uncracked part of the slab and the surrounding soil may be described by parallel springs. Spring and mass of this part of the model have to take into account the cracking of the upper slab which leads to a nonlinear characteristic of the spring. In addition the location of the loading in relation to the duct has to be considered. The duct may be described by a beam on elastic foundation which is loaded locally. From this model representative mass and spring have to be determined

  13. The energy-saving anaerobic baffled reactor membrane bioreactor (EABR-MBR) system for recycling wastewater from a high-rise building.

    Science.gov (United States)

    Ratanatamskul, Chavalit; Charoenphol, Chakraphan

    2015-01-01

    A novel energy-saving anaerobic baffled reactor-membrane bioreactor (EABR-MBR) system has been developed as a compact biological treatment system for reuse of water from a high-rise building. The anaerobic baffled reactor (ABR) compartment had five baffles and served as the anaerobic degradation zone, followed by the aerobic MBR compartment. The total operating hydraulic retention time (HRT) of the EABR-MBR system was 3 hours (2 hours for ABR compartment and very short HRT of 1 hour for aerobic MBR compartment). The wastewater came from the Charoen Wisawakam building. The results showed that treated effluent quality was quite good and highly promising for water reuse purposes. The average flux of the membrane was kept at 30 l/(m2h). The EABR-MBR system could remove chemical oxygen demand, total nitrogen and total phosphorus from building wastewater by more than 90%. Moreover, it was found that phosphorus concentration was rising in the ABR compartment due to the phosphorus release phenomenon, and then the concentration decreased rapidly in the aerobic MBR compartment due to the phosphorus uptake phenomenon. This implies that phosphorus-accumulating organisms inside the EABR-MBR system are responsible for biological phosphorus removal. The research suggests that the EABR-MBR system can be a promising system for water reuse and reclamation for high-rise building application in the near future.

  14. A fundamental approach to specify thermal and pressure loadings on containment buildings of sodium cooled fast reactors during a core disruptive accident

    International Nuclear Information System (INIS)

    Velusamy, K.; Chellapandi, P.; Satpathy, K.; Verma, Neeraj; Raviprasan, G.R.; Rajendrakumar, M.; Chetal, S.C.

    2011-01-01

    Highlights: → An approach to quantify thermal and pressure loadings on RCB is presented. → Scaling laws to determine sodium release from water experiments are proposed. → Potential of in-vessel sodium fire after a CDA is assessed. → The proposed approach is applied to Indian Prototype Fast Breeder Reactor. - Abstract: Reactor Containment Building (RCB) is the ultimate barrier to the environment against activity release in any nuclear power plant. It has to be designed to withstand both positive and negative pressures that are credible. Core Disruptive Accident (CDA) is an important event that specifies the design basis for RCB in sodium cooled fast reactors. In this paper, a fundamental approach towards quantification of thermal and pressure loadings on RCB during a CDA, has been described. Mathematical models have been derived from fundamental conservation principles towards determination of sodium release during a CDA, subsequent sodium fire inside RCB, building up of positive and negative pressures inside RCB, potential of in-vessel sodium fire due to failed seals and temperature evolution in RCB walls during extended period of containment isolation. Various heating sources for RCB air and RCB wall and their potential have been identified. Scaling laws for conducting CDA experiments in small-scale water models by chemical explosives and the rule for extrapolation of water leak to quantify sodium leak in reactor are proposed. Validation of the proposed models and experimental simulation rules has been demonstrated by applying them to Indian prototype fast breeder reactor. Finally, it is demonstrated that in-vessel sodium fire potential is very weak and no special containment cooling system is essential.

  15. Mobile means for the monitoring of atmospheric contamination in a reactor building; Moyens mobiles de surveillance de la contamination atmospherique en BR

    Energy Technology Data Exchange (ETDEWEB)

    Marques, S.; Lestang, M. [EDF/DPN/GPRE/IRP, 93 - Saint Denis (France)

    2009-07-01

    After having evoked the context and challenges of contamination monitoring when exploiting nuclear reactors, the authors discuss the representativeness of the atmospheric contamination measurement as it depends on the different physicochemical forms of radionuclides present in the circuits. They indicate the different gaseous or aerosol radioactive elements which are monitored within EDF installations. They discuss the incorporation of monitoring means at the installation design level, briefly present the use of beacons inside and outside the reactor building. They describe how monitoring is organized on the basis of alert threshold adjustments: an investigation threshold and an evacuation threshold. They discuss the beacon (or sensor) selection and indicate recommendations for their implementation for optimization purposes. They indicate where these beacons are installed and evoke the experimentation of networked mobile beacons with data remote transmission

  16. Investigation of conditions inside the reactor building annulus of a PWR plant of KONVOI type in case of severe accidents with increased containment leakages

    International Nuclear Information System (INIS)

    Bakalov, Ivan; Sonnenkalb, Martin

    2018-01-01

    Improvements of the implemented severe accident management (SAM) concepts have been done in all operating German NPPs after the Fukushima Daiichi accidents following recommendations of the German Reactor Safety Commission (RSK) and as a result of the stress test being performed. The efficiency of newly developed severe accident management guidelines (SAMG) for a PWR KONVOI reference plant related to the mitigation of challenging conditions inside the reactor building (RB) annulus due to increased containment leakages during severe accidents have been assessed. Based on two representative severe accident scenarios the releases of both hydrogen and radionuclides into the RB annulus have been predicted with different boundary conditions. The accident scenarios have been analysed without and with the impact of several SAM measures (already planned or proposed in addition), which turned out to be efficient to mitigate the consequences. The work was done within the frame of a research project financially supported by the Federal Ministry BMUB.

  17. 324 Building B-Cell Pressurized Water Reactor Spent Fuel Packaging & Shipment RL Readiness Assessment Final Report [SEC 1 Thru 3

    Energy Technology Data Exchange (ETDEWEB)

    HUMPHREYS, D C

    2002-08-01

    A parallel readiness assessment (RA) was conducted by independent Fluor Hanford (FH) and U. S. Department of Energy, Richland Operations Office (RL) team to verify that an adequate state of readiness had been achieved for activities associated with the packaging and shipping of pressurized water reactor fuel assemblies from B-Cell in the 324 Building to the interim storage area at the Canister Storage Building in the 200 Area. The RL review was conducted in parallel with the FH review in accordance with the Joint RL/FH Implementation Plan (Appendix B). The RL RA Team members were assigned a FH RA Team counterpart for the review. With this one-on-one approach, the RL RA Team was able to assess the FH Team's performance, competence, and adherence to the implementation plan and evaluate the level of facility readiness. The RL RA Team agrees with the FH determination that startup of the 324 Building B-Cell pressurized water reactor spent nuclear fuel packaging and shipping operations can safely proceed, pending completion of the identified pre-start items in the FH final report (see Appendix A), completion of the manageable list of open items included in the facility's declaration of readiness, and execution of the startup plan to operations.

  18. Thermal and stress analyses of the reactor pressure vessel lower head of the Three Mile Island Unit 2

    International Nuclear Information System (INIS)

    Hashimoto, K.; Onizawa, K.; Kurihara, R.; Kawasaki, S.; Soda, K.

    1992-01-01

    Thermal and stress analyses were performed using the finite element analysis code ABAQUS to clarify the factors which caused tears in the stainless steel liner of the reactor pressure vessel lower head of the Three Mile Island Unit 2 (TMI-2) reactor pressure vessel during the accident on 28 March 1979. The present analyses covered the events which occurred after approximately 20 tons of molten core material were relocated to the lower head of the reactor pressure vessel. They showed that the tensile stress was highest in the case where the relocated core material consisting of homogeneous UO 2 debris was assumed to attack the lower head and the debris was then quenched. The peak tensile stress was in the vicinity of the welded zone of the penetration nozzle. This result agrees with the findings from the examination of the TMI-2 reactor pressure vessel that major tears in the stainless steel liner were observed around two penetration nozzles of the lower head. (author)

  19. Neutron dosimetry in the Three-Mile Island Unit 2 reactor cavity with solid-state track recorders

    International Nuclear Information System (INIS)

    Gold, R.; Roberts, J.H.; Ruddy, F.H.; Preston, C.C.; McElroy, W.N.; Rao, S.V.; Greenborg, J.; Fricke, V.R.

    1985-01-01

    Solid-state track recorder (SSTR) neutron dosimetry has been conducted in the Three-Mile Island Unit 2 (TMI-2) reactor cavity (i.e., the annular gap between the pressure vessel and the biological shield) for nondestructive assessment of the fuel distribution. Two axial stringers were deployed in the annular gap with 17 SSTR dosimeters located on each stringer. SSTR experimental results reveal that neutron streaming, upward from the bottom of the reactor cavity region, dominates the observed neutron intensity. These absolute thermal neutron flux observations are consistent with the presence of a significant amount of fuel debris lying at the bottom of the reactor vessel. A conservative lower bound estimated from these SSTR data implies that at least 2 tonnes of fuel, which is roughly 4 fuel assemblies, is lying at the bottom of the vessel. The existence of significant neutron streaming also explains the high count rate observed with the source range monitors (SRMs) that are located in the TMI-2 reactor cavity

  20. Dynamic calculations of a PWR - reactor building for different soil parameters for the safe shutdown earthquake and explosion pressure wave load cases

    International Nuclear Information System (INIS)

    Brandt, K.; Krutzik, N.; Kaiser, A.

    1982-01-01

    For different dynamic soil properties and soil dampings - ranging from very soft to very rigid soil parameters - time histoires of displacements and accelerations as well as response spectra are calculated for several floors for the reactor building of a nuclear power plant using a finite element shell model. As regards the loadcase safety earthquake the computations are carried out for four different soil properties, and the response spectra of different floors are compared. In the loadcase exterior explosion, results for three different soils are obtained. All results are discussed and explained extensively. (Author) [pt

  1. Molten Salt Reactor Experiment Facility (Building 7503) standards/requirements identification document adherence assessment plan at Oak Ridge National Laboratory, Oak Ridge, Tennessee

    International Nuclear Information System (INIS)

    1996-02-01

    This is the Phase 2 (adherence) assessment plan for the Building 7503 Molten Salt Reactor Experiment (MSRE) Facility standards/requirements identification document (S/RID). This document outlines the activities to be conducted from FY 1996 through FY 1998 to ensure that the standards and requirements identified in the MSRE S/RID are being implemented properly. This plan is required in accordance with the Department of Energy Implementation Plan for Defense Nuclear Facilities Safety Board Recommendation 90-2, November 9, 1994, Attachment 1A. This plan addresses the major aspects of the adherence assessment and will be consistent with Energy Systems procedure QA-2. 7 ''Surveillances.''

  2. The development of the nuclear physics in Latvia II. The building of the Research Nuclear Reactor IRT

    International Nuclear Information System (INIS)

    Ulmanis, U.

    2004-01-01

    Nuclear research reactor IRT of the Academy of Sciences was built near Riga in Salaspils. IRT is pool aqueous - aqueous reactor with nuclear fuel U-235 contained elements, located in the core at a depth of ∼ 7 m under distilled water. Ten horizontal and 10-15 vertical experimental channels are employed in experimental research with the use of neutron fluxes. For the research with gamma rays is constructed radiation loop facility with liquid In-Ga-SN solid solution as intensive gamma-ray sources. Main activities of IRT are to conduct research in nuclear spectroscopy, neutron activation analysis, neutron diffraction and radiation physics, chemistry and biology. (authors)

  3. LEAF: a computer program to calculate fission product release from a reactor containment building for arbitrary radioactive decay chains

    International Nuclear Information System (INIS)

    Lee, C.E.; Apperson, C.E. Jr.; Foley, J.E.

    1976-10-01

    The report describes an analytic containment building model that is used for calculating the leakage into the environment of each isotope of an arbitrary radioactive decay chain. The model accounts for the source, the buildup, the decay, the cleanup, and the leakage of isotopes that are gas-borne inside the containment building

  4. LEAF: a computer program to calculate fission product release from a reactor containment building for arbitrary radioactive decay chains

    Energy Technology Data Exchange (ETDEWEB)

    Lee, C.E.; Apperson, C.E. Jr.; Foley, J.E.

    1976-10-01

    The report describes an analytic containment building model that is used for calculating the leakage into the environment of each isotope of an arbitrary radioactive decay chain. The model accounts for the source, the buildup, the decay, the cleanup, and the leakage of isotopes that are gas-borne inside the containment building.

  5. What we miss in order to be able to design and build a commercially viable fusion reactor

    Energy Technology Data Exchange (ETDEWEB)

    Andreani, R. [ENEA, Centro Ricerche Frascati, RM (Italy). Dipt. Energia

    1999-07-01

    The paper considers in a critical way the different areas in which work is required to provide sufficient information in view of designing a reliable and attractive fusion reactor. [Italian] Il rapporto considera in modo critico le differenti aree nelle quali si richiede ulteriore lavoro per fornire informazioni al fine di progettare un reattore a fusione affidabile ed economicamente competitivo.

  6. What we miss in order to be able to design and build a commercially viable fusion reactor

    International Nuclear Information System (INIS)

    Andreani, R.

    1999-01-01

    The paper considers in a critical way the different areas in which work is required to provide sufficient information in view of designing a reliable and attractive fusion reactor. Four main areas of activity are considered: physics, technology, engineering, safety. In physics the trend is positive towards a better understanding of suitable plasma regimes to be confirmed through further experimentation on the operating machines. Engineering has already proven itself by the design and construction of a number of experimental machines. In addition a large data base obtained from design and operation of fission reactors is available. Safety is reaching very satisfactory results in the analysis of the impact of fusion on man and the environment. Where it is still a large unsolved problem is concerning materials capable of standing the harsh fusion environment for an adequate number of years. An intense neutron source is needed in order to allow the necessary developments [it

  7. Safety of operations in the manufacture of driver fuel for the first charge of the Dragon Reactor and modifications to the safety document for the Dragon Fuel Element Production Building

    International Nuclear Information System (INIS)

    Beutler, H.; Cross, J.; Flamm, J.

    1965-01-01

    The manufacture of the zirconium containing 'driver' fuel and fuel elements for the First Charge of the Dragon Reactor Experiment has been completed without incident. This is a report on the safety of operations in the Dragon Fuel Element Production Building during an approximately six month period when the 'driver' fuel was manufactured and 25 elements containing this fuel were assembled and exported to the Reactor Building. The opportunity is taken to bring the Safety Document up-to-date and to report on any significant operational failures of equipment. (author)

  8. Streamlined Approach for Environmental Restoration Plan for Corrective Action Unit 113: Reactor Maintenance, Assembly, and Disassembly Building Nevada Test Site, Nevada

    International Nuclear Information System (INIS)

    Smith, J. L.

    2001-01-01

    This Streamlined Approach for Environmental Restoration (SAFER) Plan addresses the action necessary for the closure in place of Corrective Action Unit (CAU) 113 Area 25 Reactor Maintenance, Assembly, and Disassembly Facility (R-MAD). CAU 113 is currently listed in Appendix III of the Federal Facility Agreement and Consent Order (FFACO) (NDEP, 1996). The CAU is located in Area 25 of the Nevada Test Site (NTS) and consists of Corrective Action Site (CAS) 25-04-01, R-MAD Facility (Figures 1-2). This plan provides the methodology for closure in place of CAU 113. The site contains radiologically impacted and hazardous material. Based on preassessment field work, there is sufficient process knowledge to close in place CAU 113 using the SAFER process. At a future date when funding becomes available, the R-MAD Building (25-3110) will be demolished and inaccessible radiologic waste will be properly disposed in the Area 3 Radiological Waste Management Site (RWMS)

  9. Streamlined Approach for Environmental Restoration Plan for Corrective Action Unit 113: Reactor Maintenance, Assembly, and Disassembly Building Nevada Test Site, Nevada

    Energy Technology Data Exchange (ETDEWEB)

    J. L. Smith

    2001-01-01

    This Streamlined Approach for Environmental Restoration (SAFER) Plan addresses the action necessary for the closure in place of Corrective Action Unit (CAU) 113 Area 25 Reactor Maintenance, Assembly, and Disassembly Facility (R-MAD). CAU 113 is currently listed in Appendix III of the Federal Facility Agreement and Consent Order (FFACO) (NDEP, 1996). The CAU is located in Area 25 of the Nevada Test Site (NTS) and consists of Corrective Action Site (CAS) 25-04-01, R-MAD Facility (Figures 1-2). This plan provides the methodology for closure in place of CAU 113. The site contains radiologically impacted and hazardous material. Based on preassessment field work, there is sufficient process knowledge to close in place CAU 113 using the SAFER process. At a future date when funding becomes available, the R-MAD Building (25-3110) will be demolished and inaccessible radiologic waste will be properly disposed in the Area 3 Radiological Waste Management Site (RWMS).

  10. Confirmation test on the dynamic interaction between a model reactor-building foundation and ground in the Sendai Nuclear Power Station

    International Nuclear Information System (INIS)

    Umezu, Hideo; Kisaki, Noboru; Shiota, Mutsumi

    1982-01-01

    On the site of unit 2 (planned) in the Sendai Nuclear Power Station, a model reactor-building foundation of reinforced concrete with diameter of 12 m and height of 5 m was installed. With a vibration generator, its forced vibration tests were carried out in October to December, 1980. Valuable data were able to be obtained on the dynamic interaction between the model foundation and the ground, and also the outlook for the application of theories in hard base rock was obtained. (1) The resonance frequency of the model foundation in horizontal vibration was 35 Hz in both NS and EW directions. (2) Remarkable difference was not observed in the horizontal vibration behavior between NS and EW directions, so that there is not anisotropy in the ground. (3) The model foundation was deformed nearly as a rigid body. (J.P.N.)

  11. Long-Term Behaviors of the OPC Concrete with Fly-ash and Type V Concrete Applied on Reactor Containment Building

    International Nuclear Information System (INIS)

    Yoon, Eui Sik; Lee, Hee Taik; Paek, Yong Lak; Park, Young Soo

    2010-01-01

    The prestressed concrete has been used extensively in the construction of Reactor Containment Buildings (RCBs) in Korea in order to strengthen the RCBs and at the same time, prevent the release of radiation due to the Design Basis Accident and Design Basis Earthquake. It is well known that the prestressed concrete loses its prestressing force over the age, and the shrinkage and creep of the concrete significantly contributes to these long term prestressing losses. In this study, an evaluations of long term behaviors of the concrete such as creep and shrinkage have been performed for two types of concretes : Ordinary Portland Cement containing fly-ash used for the Shin- Kori 1 and 2 NPP and Type V cement used for the Ul- Chin 5 and 6 NPP

  12. Assessments of the probabilities of aircraft impact with the Sandia Pulsed Reactor and Building 836, Sandia Laboratories, Albuquerque

    International Nuclear Information System (INIS)

    Biringer, B.E.

    1976-11-01

    This report documents a study of the annual probabilities of aircraft impact with the Sandia Pulsed Reactor (SPR) and Bldg. 836 at Sandia Laboratories, Albuquerque. The probability of aircraft impact into each structure was estimated using total yearly operations, effective structure area, structure location relative to air activity, and accident rate per kilometer. The estimated probability for an aircraft impact with SPR is 1.1 x 10 -4 per year; the estimated probability for impact with Bldg. 836 is 1.0 x 10 -3 per year

  13. Combination of a building condenser with H{sub 2}-recombiner elements in light water reactors; Kombination eines Gebaeudekondensators mit H{sub 2}-Rekombinatorelementen in Leichtwasserreaktoren

    Energy Technology Data Exchange (ETDEWEB)

    Kelm, Stephan

    2010-06-17

    In innovative reactor concepts of generation III and III+, passive safety systems are applied in order to ensure the integrity of the physical barriers against the release of fission products to the environment in the course of accidents. The containment represents the final barrier. An early filtered release due to an unacceptable pressure increase therefore has to be excluded or at least delayed. For this purpose, building condensers are envisaged in modern containment concepts in order to condense the steam released during an accident and thus reduce pressure. Experimental investigations on this concept conducted at various facilities showed a significant trip of the condenser effectivity in case of beyond-design-basis accidents with hydrogen release into the containment. In order to reduce the related significant pressure rise, an innovative condenser concept with integrated catalyst elements for recombining the hydrogen with the atmospheric oxygen present it proposed and investigated. The CFD (computational fluid dynamics) models developed in the present work form the basis for numerical simulation of the complex interaction between catalytic reaction, wall condensation and buoyancy-driven flow inside an innovative, combined catalyst-condenser bundle. The high-resolution CFD simulation allows for the first time detailed insight into the flow and transport phenomena inside the tube bundle, verifies the conceptual idea and thus serves for the enhancement of safety margins of future reactor systems. (orig.)

  14. Replacement of the feedwater pipe system in reactor building outside containment at the nuclear power plant Philippsburg; Austausch der Speisewasserleitung im Reaktorgebaeude ausserhalb SHB im Kernkraftwerk Philippsburg I

    Energy Technology Data Exchange (ETDEWEB)

    Kessler, A. [Energie-Versorgung Schwaben AG, Stuttgart (Germany); Labes, M. [Siemens AG Unternehmensbereich KWU, Offenbach am Main (Germany); Schwenk, B. [Kernkraftwerk Philippsburg GmbH (Germany)

    1998-11-01

    After full replacement of the feedwater pipe system during the inspection period in 1997, combined with a modern materials, manufacturing and analysis concept, the entire pipe system of the water/steam cycle in the reactor building of KKP 1 now consists of high-toughness materials. The safety level of the entire plant has been increased by leaving aside postulation of F2 breaks in the reactor building and providing for protection against 0.1 leaks. Based on fluid-dynamic calculations for the cases of pump failure and pipe break, as well as pipe system calculations in 5 extensive calculation cycles, about 130 documents were filed for inspection and approval (excluding preliminary test documents on restraints). Points of main interest for safety analysis in this context were the optimised closing performance of the 3rd check valves and the integrity of the nozzle region at the RPV. (oirg./CB) [Deutsch] Durch den Restaustausch der Speisewasserleitungen in der Revision 1997, verbunden mit einem modernen Werkstoff-, Fertigungs- und Nachweiskonzept, sind im Reaktorgebaeude von KKP 1 in den Hauptleitungen des Wasser-Dampf-Kreislaufes nur noch hochzaehe Werkstoffe eingesetzt. Durch den Verzicht auf das Postulat von 2F-Bruechen im Reaktorgebaeude und durch die Auslegung gegen 0,1F-Lecks wird das Sicherheitsniveau der Anlage insgesamt gesteigert. Ausgehend von fluiddynamischen Berechnungen fuer Pumpenausfall und Rohrbruch sowie Rohrsystem-Berechnungen in 5 umfangreichen Berechnungskreisen wurden fuer die Genehmigung und Begutachtung ca. 130 Unterlagen (ohne Halterungs-Vorpruefunterlagen) eingereicht und vom Gutachter geprueft. Schwerpunkte der Nachweisfuehrung waren die Optimierung des Schliessverhaltens der 3. Rueckschlagarmaturen sowie der Integritaetsnachweis des RDB-Anschlusses. (orig./MM)

  15. A Stencil to Develop an EOG for a Research Reactor

    International Nuclear Information System (INIS)

    Kim, Hyeonil; Yum, Soo-been; Jung, Youn-Gyu; Lee, Byeonghee; Jun, In Sub; Park, Su-Ki

    2016-01-01

    Following the Three Mile Island Unit 2 (TMI-2) accident, the United States Nuclear Regulatory Commission (NRC) established the sets of requirements addressing their objective to improve the quality of operational information for dealing with emergency events in nuclear power plants. The Emergency Operating Guidelines (EOG) should be presented to provide technical information to prepare reactor-specific Emergency Operating Procedures (EOP) which cover operation during emergency events. EOG must provide guidance for both classes of emergencies. Thus, when a reactor trip occurs or should occur, the operators can refer to guidance which will provide a safe response whether or not a symptom set is identified: EOG written to treat specific symptoms are called event-based recovery guidelines (ERG); the EOG which provides guidance for undiagnosed events for which a reactor trip is required is called the Symptom-based Recovery Guidelines (SRGs). For helping to design and develop a set of EOG/EOP, a stencil in a sheet was proposed as an easy and intuitive tool to gather information of a research reactor related to safety/safety-related functions. With concept of safety functions, the stencil will be able to give a strong but very easy, straightforward, systematic, comprehensive tool to analyze the architecture of the reactor, to cover the whole SSC information

  16. What we miss in order to be able to design and build a commercially viable fusion reactor

    Energy Technology Data Exchange (ETDEWEB)

    Andreani, R. [ENEA, Centro Ricerche Frascati, Rome (Italy). Dipt. Energia

    1999-07-01

    The paper considers in a critical way the different areas in which work is required to provide sufficient information in view of designing a reliable and attractive fusion reactor. Four main areas of activity are considered: physics, technology, engineering, safety. In physics the trend is positive towards a better understanding of suitable plasma regimes to be confirmed through further experimentation on the operating machines. Engineering has already proven itself by the design and construction of a number of experimental machines. In addition a large data base obtained from design and operation of fission reactors is available. Safety is reaching very satisfactory results in the analysis of the impact of fusion on man and the environment. Where it is still a large unsolved problem is concerning materials capable of standing the harsh fusion environment for an adequate number of years. An intense neutron source is needed in order to allow the necessary developments. [Italian] Il rapporto considera in modo critico le differenti aree nelle quali si richiede ulteriore lavoro per fornire informazini sufficienti al fine di progettare un reattore a fusione affidabile ed economicamente competitivo. Vengono considerate quattro aree principali di attivita': fisica, tecnologia, ingegneria, sicurezza. Nella fisica, vi e' una positiva tendenza verso una migliore comprensione di regimi di plasma favorevoli da confermare attraverso ulteriore sperimentazione sulle macchine funzionanti. L' ingegneria ha gia' dato dimostrazione di se' col progetto e la costruzione di un notevole numero di macchine sperimentali. In aggiunta e' disponibile un gran numero di dati ottenuti dalla progettazione, realizzazione e funzionamento dei reattori a fissione. La sicurezza sta raggiungendo risultati molto soddisfacenti nell'analisi dell'impatto della fusione sull'uomo e sull'ambiente. Un grosso problema tuttora irresoluto e' quello dei

  17. Nuclear Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Hogerton, John

    1964-01-01

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

  18. Determination of the exposition rapidity in the level 49.90 of the reactor building for the decrease in the water level of the spent fuel pool

    International Nuclear Information System (INIS)

    Mijangos D, Z. E.; Herrera H, S. F.; Cruz G, M. A.; Amador C, C.

    2014-10-01

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

  19. Regulations for RA reactor operation

    International Nuclear Information System (INIS)

    1980-09-01

    Regulations for RA reactor operation are written in accordance with the legal regulations defined by the Law about radiation protection and related legal acts, as well as technical standards according to the IAEA recommendations. The contents of this book include: fundamental data about the reactor; legal regulations for reactor operation; organizational scheme for reactor operation; general and detailed instructions for operation, behaviour in the reactor building, performing experiments; operating rules for operation under steady state and accidental conditions [sr

  20. Developments in radiological protection of the environment and a commentary on its implications for new build of nuclear reactors

    International Nuclear Information System (INIS)

    Brownless, George; Lazo, Ted

    2008-01-01

    Full text: Development of radiological protection of non-human biota continues to be the focus of much interest and differing views amongst the radiological protection community. To nurture discussion of these developments, the Nuclear Energy Agency's Committee on Radiological Protection and Public Health is taking the lead in organising coverage of a spectrum of views on the topic at the International Conference on Radioecology and Radioactivity in the Environment (Bergen, 2008), with the aim of assisting the international community to construct a consensual, fit-for-purpose approach. To support discussion of these views, the session will also include scientific presentations and reports from implementers. This paper will report on these developments, based principally on the session at the Conference but also other activities in which NEA participates, to provide an up-to-date summary in this area including progress in developing understanding of how radiological protection of the environment will be implemented for the three exposure situations - planned, existing and emergency - set out in the new ICRP general recommendations. Furthermore, given the NEA's mandate across the civil nuclear energy field, the paper will give a commentary on how developments in radiological protection of the environment may interplay with new build of nuclear power plants. (author)

  1. Nuclear reactor container

    International Nuclear Information System (INIS)

    Takahashi, Hiroyuki.

    1987-01-01

    Purpose: To improve the earthquake proofness and also increase the safety to a nuclear reactor container by preventing bucklings upon earthquake. Constitution: A device for absorbing the deformation exerted from nuclear reactor buildings is disposed to a suppression chamber constituting a reactor container. When a nclear power plant encounters earthquakes, the entire reactor buildings are shaken and deformations of buildings are transmitted by way of building shell walls to a container and the forcive deforming forces are absorbed in the deformation absorbing device. That is, bellows are formed at the base of the container, which are deformed by the deforming forces to absorb the forcive deforming amount to moderate the stresses resulted to the suppression chamber. Thus, the rigidity to the bending of the container can be reduced and allowable displacement to the bucklings can be increased to prevent the buckling, by which earthquake proofness is improved and the safety is increased. (Kamimura, M.)

  2. Technology, safety and costs of decommissioning a reference pressurized water reactor power station: Technical support for decommissioning matters related to preparation of the final decommissioning rule

    Energy Technology Data Exchange (ETDEWEB)

    Konzek, G.J.; Smith, R.I.

    1988-07-01

    Preparation of the final Decommissioning Rule by the Nuclear Regulatory Commission (NRC) staff has been assisted by Pacific Northwest Laboratory (PNL) staff familiar with decommissioning matters. These efforts have included updating previous cost estimates developed during the series of studies on conceptually decommissioning reference licensed nuclear facilities for inclusion in the Final Generic Environmental Impact Statement (FGEIS) on decommissioning; documenting the cost updates; evaluating the cost and dose impacts of post-TMI-2 backfits on decommissioning; developing a revised scaling formula for estimating decommissioning costs for reactor plants different in size from the reference pressurized water reactor (PWR) described in the earlier study; defining a formula for adjusting current cost estimates to reflect future escalation in labor, materials, and waste disposal costs; and completing a study of recent PWR steam generator replacements to determine realistic estimates for time, costs and doses associated with steam generator removal during decommissioning. This report presents the results of recent PNL studies to provide supporting information in four areas concerning decommissioning of the reference PWR: updating the previous cost estimates to January 1986 dollars; assessing the cost and dose impacts of post-TMI-2 backfits; assessing the cost and dose impacts of recent steam generator replacements; and developing a scaling formula for plants different in size than the reference plant and an escalation formula for adjusting current cost estimates for future escalation.

  3. Technology, safety and costs of decommissioning a reference boiling water reactor power station: Technical support for decommissioning matters related to preparation of the final decommissioning rule

    Energy Technology Data Exchange (ETDEWEB)

    Konzek, G.J.; Smith, R.I.

    1988-07-01

    Preparation of the final Decommissioning Rule by the Nuclear Regulatory Commission (NRC) staff has been assisted by Pacific Northwest Laboratory (PNL) staff familiar with decommissioning matters. These efforts have included updating previous cost estimates developed during the series of studies of conceptually decommissioning reference licensed nuclear facilities for inclusion in the Final Generic Environmental Impact Statement (FGEIS) on decommissioning; documenting the cost updates; evaluating the cost and dose impacts of post-TMI-2 backfits on decommissioning; developing a revised scaling formula for estimating decommissioning costs for reactor plants different in size from the reference boiling water reactor (BWR) described in the earlier study; and defining a formula for adjusting current cost estimates to reflect future escalation in labor, materials, and waste disposal costs. This report presents the results of recent PNL studies to provide supporting information in three areas concerning decommissioning of the reference BWR: updating the previous cost estimates to January 1986 dollars; assessing the cost and dose impacts of post-TMI-2 backfits; and developing a scaling formula for plants different in size than the reference plant and an escalation formula for adjusting current cost estimates for future escalation.

  4. Technology, safety and costs of decommissioning a refernce boiling water reactor power station: Technical support for decommissioning matters related to preparation of the final decommissioning rule

    International Nuclear Information System (INIS)

    Konzek, G.J.; Smith, R.I.

    1988-07-01

    Preparation of the final Decommissioning Rule by the Nuclear Regulatory Commission (NRC) staff has been assisted by Pacific Northwest Laboratory (PNL) staff familiar with decommissioning matters. These efforts have included updating previous cost estimates developed during the series of studies of conceptually decommissioning reference licensed nuclear facilities for inclusion in the Final Generic Environmental Impact Statement (FGEIS) on decommissioning; documenting the cost updates; evaluating the cost and dose impacts of post-TMI-2 backfits on decommissioning; developing a revised scaling formula for estimating decommissioning costs for reactor plants different in size from the reference boiling water reactor (BWR) described in the earlier study; and defining a formula for adjusting current cost estimates to reflect future escalation in labor, materials, and waste disposal costs. This report presents the results of recent PNL studies to provide supporting information in three areas concerning decommissioning of the reference BWR: updating the previous cost estimates to January 1986 dollars; assessing the cost and dose impacts of post-TMI-2 backfits; and developing a scaling formula for plants different in size than the reference plant and an escalation formula for adjusting current cost estimates for future escalation

  5. Technology, safety and costs of decommissioning a reference pressurized water reactor power station: Technical support for decommissioning matters related to preparation of the final decommissioning rule

    International Nuclear Information System (INIS)

    Konzek, G.J.; Smith, R.I.

    1988-07-01

    Preparation of the final Decommissioning Rule by the Nuclear Regulatory Commission (NRC) staff has been assisted by Pacific Northwest Laboratory (PNL) staff familiar with decommissioning matters. These efforts have included updating previous cost estimates developed during the series of studies on conceptually decommissioning reference licensed nuclear facilities for inclusion in the Final Generic Environmental Impact Statement (FGEIS) on decommissioning; documenting the cost updates; evaluating the cost and dose impacts of post-TMI-2 backfits on decommissioning; developing a revised scaling formula for estimating decommissioning costs for reactor plants different in size from the reference pressurized water reactor (PWR) described in the earlier study; defining a formula for adjusting current cost estimates to reflect future escalation in labor, materials, and waste disposal costs; and completing a study of recent PWR steam generator replacements to determine realistic estimates for time, costs and doses associated with steam generator removal during decommissioning. This report presents the results of recent PNL studies to provide supporting information in four areas concerning decommissioning of the reference PWR: updating the previous cost estimates to January 1986 dollars; assessing the cost and dose impacts of post-TMI-2 backfits; assessing the cost and dose impacts of recent steam generator replacements; and developing a scaling formula for plants different in size than the reference plant and an escalation formula for adjusting current cost estimates for future escalation

  6. Radio-contaminant behaviour in the cover-gas space and the containment building of a sodium-cooled fast reactor in accident conditions

    International Nuclear Information System (INIS)

    Mathe, Emmanuel

    2014-01-01

    In the context of the Generation IV initiative, the consequences of a severe-accident (SA) in a sodium-cooled fast reactor must be studied. A SFR (Sodium cooled Fast Reactor) severe accident involves the disruption of the core by super-criticality involving the destruction of a certain number of fuel assemblies. Subsequently the interaction between hot fuel and liquid sodium can lead to a vapor explosion which could create a breach in the primary system. Some contaminated liquid sodium would thus be ejected into the containment building. In this situation, the evaluation of potential releases to the environment (the source term) must forecast the quantity and the chemical speciation of the radio-contaminants likely to be released from the containment building. One critical risk of a SA is the production of contaminated aerosols in the containment building by spray ejection of primary-system sodium. Being pyrophoric, the sodium droplets react with oxygen first oxidizing then burning, with significant heat of combustion. As well as evaluating the consequences of a pressure rise inside the containment, the evolution of the sodium must be assessed since not only is it activated and contaminated but, in oxide form, very toxic. Ultimately, the aerosols are the main radiological risk acting as the vector for radionuclide transport to the environment in the event of a problem with the confinement. These aerosols could evolve and interact with the FP (Fissile Products) and these interactions could modify the physical and chemical nature of the PF. We model a large part of the events that occur during a SA inside a SFR from the sodium spray fire to the reaction between sodium aerosols and PF (iodine). At first, we develop a numerical model (NATRAC) that simulates the sodium spray fire, calculates the temperature and the pressure inside the containment as well as the mass of aerosols produced during this kind of fire. The simulation has been validated with different

  7. One piece reactor removal

    International Nuclear Information System (INIS)

    Chia, Wei-Min; Wang, Song-Feng

    1993-01-01

    The strategy of Taiwan Research Reactor Renewal plan is to remove the old reactor block with One Piece Reactor Removal (OPRR) method for installing a new research reactor in original building. In this paper, the engineering design of each transportation works including the work method, the major equipments, the design policy and design criteria is described and discussed. In addition, to ensure the reactor block is safety transported for storage and to guarantee the integrity of reactor base mat is maintained for new reactor, operation safety is drawn special attention, particularly under seismic condition, to warrant safe operation of OPRR. ALARA principle and Below Regulatory Concern (BRC) practice were also incorporated in the planning to minimize the collective dose and the total amount of radioactive wastes. All these activities are introduced in this paper. (J.P.N.)

  8. Nuclear future: thinking for building. Proceedings of the 12. Brazilian national meeting on reactor physics and thermal hydraulics; 8. General congress on nuclear energy; 5. Brazilian national meeting on nuclear applications

    International Nuclear Information System (INIS)

    2000-01-01

    These proceedings, for the first time, present jointly the 12. Brazilian national meeting on reactor physics and thermal hydraulics (12 ENFIR), 8. General congress on nuclear energy (8. CGEN), and 5. Brazilian national meeting on nuclear applications (5. ENAN). The main theme of discussion was: 'Nuclear Future: thinking for building'. The papers have analysed the progresses of peaceful utilization of nuclear technology and its forecasting for the beginning of the new millennium. The construction of Angra-3 nuclear power plant have been discussed

  9. Reactor container

    International Nuclear Information System (INIS)

    Otsuka, Hiroaki; Yoshida, Takashi.

    1979-01-01

    Purpose: To prevent rain water falling along the outer wall of the container during the construction work of an atomic power plant from making ingress into the inner part of a reactor container through a large size material carry-in port. Constitution: A weir for preventing the ingress of rain water is provided on the border between the foot floor of a large material carry-in port provided on the side surface at the bottom part of the reactor container and the floor surface of the building. This weir is of a semi-circular plate shape, and formed so that the lower semi-circular part of the carry-in port is tightly closed. (Kamimura, M.)

  10. A study on the late core melt progression in pressurized water reactor severe accidents

    International Nuclear Information System (INIS)

    Park, Jae Hong; Jeun Gyoo Dong; Bang, Kwang Hyun; Park, Seh In; Lim, Jae Hyuck; Park, Seong Yong; Back, Hyung Hmm

    1998-03-01

    After TMI-2 accidents, it has been paid much attention to severe accidents beyond the design basis accidents and the research on the progress of severe accidents and mitigation and the closure of severe accidents has been actively performed. In particular, a great deal of uncertainties yet exist in the phase of late core melt progression and thus the research on this phase of severe accident progress has a key role in obtaining in severe accident mitigation and nuclear reactor safety. In the present study, physics of late core melt progression, experimental data and the major phenomenological models of computer codes are reviewed and a direction of reducing the uncertainties in the late core melt progression os proposed

  11. Fusion Reactor Materials

    International Nuclear Information System (INIS)

    Decreton, M.

    2002-01-01

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

  12. RB research reactor Safety Report

    International Nuclear Information System (INIS)

    Sotic, O.; Pesic, M.; Vranic, S.

    1979-04-01

    This RB reactor safety report is a revised and improved version of the Safety report written in 1962. It contains descriptions of: reactor building, reactor hall, control room, laboratories, reactor components, reactor control system, heavy water loop, neutron source, safety system, dosimetry system, alarm system, neutron converter, experimental channels. Safety aspects of the reactor operation include analyses of accident causes, errors during operation, measures for preventing uncontrolled activity changes, analysis of the maximum possible accident in case of different core configurations with natural uranium, slightly and highly enriched fuel; influence of possible seismic events

  13. Working with the States to Transport TMI-2 Core Debris

    International Nuclear Information System (INIS)

    Smith, T.A.; Anselmo, A.A.

    1989-01-01

    Close communications with state officials has been a key factor in success of the Three Mile Island Unit 2 core debris shipments. The U.S. Department of Energy made extensive efforts to provide state officials with schedule information, answer technical questions, and satisfy concerns. Communications started before the campaign and continued during shipments and at intervals between shipments. Those efforts led to good working relationships with the states, kept governors and other state officials informed so they could respond to public concerns, provided the opportunity to recognize and respond to specific state concerns, facilitated state inspections, and provided avenues to avoid conflict and potential litigation. Good communications and working relationships with state officials also greatly benefited the community relations effort for the campaign. (author)

  14. TMI 2 and safety culture. Lessons still to be learned

    International Nuclear Information System (INIS)

    Werdine, H.

    1998-01-01

    This paper will deal with some concerns related to the actual ''status-quo'' of the nuclear industry related to nuclear safety. It is well understood that incredible progresses have been made in this subject not only by the industries, but mainly by the International Organisations such as INPO, WANO, IAEA and others. Of course, there could be some misunderstanding and corrections here and there, and now I am talking about the WANO Mission last year in Chernobyl. But this subject is not the objective of my paper. This paper deals with a concern. I am worried with the actual status of a ''potential relative'' complacency with some very important safety principles that could be permeating throughout the nuclear industry. I will try to develop this idea and propose some actions to counteract the facts; what we did at Angra 1 and within ELETRONUCLEAR to face those challenges. (author)

  15. Robotics program development: Applicable lessons learned from TMI-2

    International Nuclear Information System (INIS)

    Schwartz, F.L.

    1989-09-01

    The nuclear industry is becoming increasingly aware of the benefits that can result from the use of robotics technology to augment plant maintenance and monitoring programs. This awareness is growing as a result of various utility successes in deploying robotics tools and realizing benefits to plant operations. These benefits include: (1) reducing occupational exposure risks associated with performing tasks in high radiation zones; (2) reducing fatigue associated with physically demanding tasks; and (3) minimizing human error associated with repetitive tasks. This EPRI-sponsored report addresses the programmatic issues faced by GPU Nuclear in developing a program that extensively employed robotics and remotely controlled equipment. Many of the lessons learned can be applied to an operating plant. This report has been prepared to present generally applicable experiences that a utility may wish to consider when developing a robotics program. Issues addressed within this report can serve as a preliminary guide to: (1) establishing equipment requirements; (2) preparing personnel to implement a robotics program; and (3) determining the impacts on the administrative aspects of operations related to procedures, licensing, and reliability. The Three Mile Island Unit 2 accident created a unique situation within industry; i.e., remotely controlled equipment provided the only viable option for accomplishing a number of recovery tasks. This impetus accelerated the development of practical robotics tooling for specific tasks. 16 refs., 35 figs., 2 tabs

  16. Community Relations for the Transport of TMI-2 Core Debris

    International Nuclear Information System (INIS)

    Smith, T.A.

    1988-01-01

    This paper describes community relations for the transport of Three Mile Island Unit 2 core debris, before and during the first two years of the campaign. The author defines community relations as interactions with groups or individuals to influence public perception. Members of Congress, state and local officials, news media, special interest groups, and private citizens are included in the definition of community. The paper discusses issues of concern to the community, level of interest generated by the transport campaign, events that kept community interest focused on the campaign, and communication techniques employed to provide the community with factual information and to generate public confidence. Finally, the paper describes lessons learned from the community relations effort. (author)

  17. Thermal reactor safety

    International Nuclear Information System (INIS)

    1980-06-01

    Information is presented concerning new trends in licensing; seismic considerations and system structural behavior; TMI-2 risk assessment and thermal hydraulics; statistical assessment of potential accidents and verification of computational methods; issues with respect to improved safety; human factors in nuclear power plant operation; diagnostics and activities in support of recovery; LOCA transient analysis; unresolved safety issues and other safety considerations; and fission product transport

  18. Thermal reactor safety

    Energy Technology Data Exchange (ETDEWEB)

    1980-06-01

    Information is presented concerning new trends in licensing; seismic considerations and system structural behavior; TMI-2 risk assessment and thermal hydraulics; statistical assessment of potential accidents and verification of computational methods; issues with respect to improved safety; human factors in nuclear power plant operation; diagnostics and activities in support of recovery; LOCA transient analysis; unresolved safety issues and other safety considerations; and fission product transport.

  19. In-vessel Retention Strategy for High Power Reactors - K-INERI Final Report (includes SBLB Test Results for Task 3 on External Reactor Vessel Cooling (ERVC) Boiling Data and CHF Enhancement Correlations)

    Energy Technology Data Exchange (ETDEWEB)

    F. B. Cheung; J. Yang; M. B. Dizon; J. Rempe

    2005-01-01

    In-vessel retention (IVR) of core melt is a key severe accident management strategy adopted by some operating nuclear power plants and proposed for some advanced light water reactors (ALWRs). If there were inadequate cooling during a reactor accident, a significant amount of core material could become molten and relocate to the lower head of the reactor vessel, as happened in the Three Mile Island Unit 2 (TMI-2) accident. If it is possible to ensure that the vessel head remains intact so that relocated core materials are retained within the vessel, the enhanced safety associated with these plants can reduce concerns about containment failure and associated risk. For example, the enhanced safety of the Westinghouse Advanced 600 MWe PWR (AP600), which relied upon External Reactor Vessel Cooling (ERVC) for IVR, resulted in the U.S. Nuclear Regulatory Commission (US NRC) approving the design without requiring certain conventional features common to existing LWRs. However, it is not clear that currently proposed external reactor vessel cooling (ERVC) without additional enhancements could provide sufficient heat removal for higher-power reactors (up to 1500 MWe). Hence, a collaborative, three-year, U.S. - Korean International Nuclear Energy Research Initiative (INERI) project was completed in which the Idaho National Engineering and Environmental Laboratory (INEEL), Seoul National University (SNU), Pennsylvania State University (PSU), and the Korea Atomic Energy Research Institute (KAERI) investigated the performance of ERVC and an in-vessel core catcher (IVCC) to determine if IVR is feasible for reactors up to 1500 MWe.

  20. Nuclear problems of power reactors safety

    International Nuclear Information System (INIS)

    Kocic, A.

    1980-01-01

    The objective of this presentation is to emphasize the conditions that would be of high importance in safety analyses concerned first of all with reactor core. It describes reactor kinetics processes in the core and build up of fission products, classification of reactor accidents related to the core, risk estimation and includes a list of importance reactor accidents

  1. Nuclear reactors

    International Nuclear Information System (INIS)

    Barre, Bertrand

    2015-10-01

    After some remarks on the nuclear fuel, on the chain reaction control, on fuel loading and unloading, this article proposes descriptions of the design, principles and operations of different types of nuclear reactors as well as comments on their presence and use in different countries: pressurized water reactors (design of the primary and secondary circuits, volume and chemistry control, backup injection circuits), boiling water reactors, heavy water reactors, graphite and boiling water reactors, graphite-gas reactors, fast breeder reactors, and fourth generation reactors (definition, fast breeding). For these last ones, six concepts are presented: sodium-cooled fast reactor, lead-cooled fast reactor, gas-cooled fast reactor, high temperature gas-cooled reactor, supercritical water-cooled reactor, and molten salt reactor

  2. Nuclear future: thinking for building. Proceedings of the 5. Brazilian national meeting on nuclear applications; 8. General congress on nuclear energy; 12. Brazilian national meeting on reactor physics and thermal hydraulics

    International Nuclear Information System (INIS)

    2000-01-01

    These proceedings, for the first time, present jointly the 12. Brazilian national meeting on reactor physics and thermal hydraulics (12. ENFIR), the 8. General congress on nuclear energy (8. CGEN), and the 5. Brazilian national meeting on nuclear applications (5. ENAN). The main theme of discussion was: 'Nuclear Future: thinking for building'. The papers have analysed the progresses of peaceful utilization of nuclear technology and its forecasting for the beginning of the new millennium. The construction of Angra-3 nuclear power plant have been discussed

  3. Earthquake proof device for nuclear power plant building

    International Nuclear Information System (INIS)

    Okada, Yasuo.

    1991-01-01

    The structure of the present invention enables three dimensional vibration proof, i.e., in horizontal and vertical directions of a reactor container building. That is, each of the reactor container building and a reactor auxiliary building is adapted as an independent structure. The periphery of the reactor container building is surrounded by the reactor auxiliary building. The reactor auxiliary building is supported against the ground by way of a horizontal vibration proof device. The reactor container building is supported against the ground by way of a three-dimensional vibration proof device that prevents vibrations in both of the horizontal directions, and the vertical directions. The reactor container building is connected to the auxiliary building by way of a vertical vibration proof device. With such a constitution, although the reactor container building is vibration proof against both of the horizontal and the vertical vibrations, the vertical vibration proofness is an extension of inherent vertical vibration period. Accordingly, the head of the building undergoes rocking vibrations. However, since the reactor container building is connected to the reactor auxiliary building, the rocking vibrations are prevented by the reactor auxiliary building. As a result, safety upon occurrence of an earthquakes can be ensured. (I.S.)

  4. RA reactor operation and maintenance in 1992, Part 1

    International Nuclear Information System (INIS)

    Sotic, O.; Cupac, S.; Sulem, B.; Zivotic, Z.; Majstorovic, D.; Tanaskovic, M.

    1992-01-01

    During 1992 Ra reactor was not in operation. All the activities were fulfilled according to the previously adopted plan. Basic activities were concerned with revitalisation of the RA reactor and maintenance of reactor components. All the reactor personnel was busy with reconstruction and renewal of the existing reactor systems and building of the new systems, maintenance of the reactor devices. Part of the staff was trained for relevant tasks and maintenance of reactor systems [sr

  5. H Reactor

    Data.gov (United States)

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

  6. Report on design and technical standard planning of vibration controlling structure on the buildings, in the Tokai Reprocessing Facility, Power Reactor and Nuclear Fuel Development Corporation

    International Nuclear Information System (INIS)

    Uryu, Mitsuru; Terada, Shuji; Shinohara, Takaharu; Yamazaki, Toshihiko; Nakayama, Kazuhiko; Kondo, Toshinari; Hosoya, Hisashi

    1997-10-01

    The Tokai reprocessing facility buildings are constituted by a lower foundation, vibration controlling layers, and upper structure. At the vibration controlling layer, a laminated rubber aiming support of the building load and extension of the eigenfrequency and a damper aiming absorption of earthquake energy are provided. Of course, the facility buildings are directly supported at the arenaceous shale (Taga Layer) of the Miocene in the Neogene confirmed to the stablest ground, as well the buildings with high vibration resistant importance in Japan. This report shows that when the vibration controlling structure is adopted for the reprocessing facility buildings where such high vibration resistance is required, reduction of input acceleration for equipments and pipings can be achieved and the earthquake resistant safety can also be maintained with sufficient tolerance and reliability. (G.K.)

  7. PARR-2: reactor description and experiments

    International Nuclear Information System (INIS)

    Wyne, M.F.; Meghji, J.H.

    1990-12-01

    PARR-2 is a miniature neutron source reactor (MNSR) research reactor has been designed at the rate of 27 kW. Reactor assembly comprises of peaking characteristics with a self limiting flux. In this report reactor description with its assembly and instrumentation control system has been explained. The reactor engineering and physics experiments which can be performed on this reactor are explained in this report. PARR-2 is fueled with HEU fuel pins which are about 90% enriched in U-235. Specific requirements for the safety of the reactor, its building and the personnel, normal instrumentation as required in an industrial environment is sufficient. (A.B.)

  8. The safety of operations in the Dragon fuel element production building during the manufacture of thorium fuel for the first charge of the Dragon Reactor experiment

    International Nuclear Information System (INIS)

    Beutler, H.; Gardham, B.; Holliday, J.

    1965-04-01

    The first charge of fuel and fuel elements for the Dragon Reactor has been completed without significant difficulty. This report covers the safety of operations during the production of the 10 thorium elements together with the final 2 driver elements needed to complete the 37 element charge. (author)

  9. Final-Independent Confirmatory Survey Report For The Reactor Building, Hot Laboratory, Primary Pump House, And Land Areas At The Plum Brook Reactor Facility, Sandusky, Ohio DCN:2036-SR-01-10

    International Nuclear Information System (INIS)

    Bailey, Erika N.

    2011-01-01

    In 1941, the War Department acquired approximately 9,000 acres of land near Sandusky, Ohio and constructed a munitions plant. The Plum Brook Ordnance Works Plant produced munitions, such as TNT, until the end of World War II. Following the war, the land remained idle until the National Advisory Committee for Aeronautics later called the National Aeronautics and Space Administration (NASA) obtained 500 acres to construct a nuclear research reactor designed to study the effects of radiation on materials used in space flight. The research reactor was put into operation in 1961 and was the first of fifteen test facilities eventually built by NASA at the Plum Brook Station. By 1963, NASA had acquired the remaining land at Plum Brook for these additional test facilities

  10. Challenges in decontamination technology

    International Nuclear Information System (INIS)

    Bengal, P.R.; Davis, C.M.

    1986-01-01

    The recovery of the Three Mile Island Unit 2 facilities has transformed decontamination from traditional ''mop-and-bucket'' cleaning into an advanced technology. The decontamination of the reactor building and auxiliary and fuel handling building required the development of new methodologies, adaptation of equipment from other fields, and the development of new equipment to meet the projects goals. Experience at TMI-2 has resulted in the development of techniques for; radiological characterization with wide angle detectors and autoradiograhic techniques for the determination of sources and depth of contamination penetration, coatings and base surface removal with ultra-high-pressure water dissolution chemicals and mechanical scabblers (paint-chipping devices), and surface decontamination with high-pressure steam vacuum extraction. This paper presents an overview of the development of decontamination technology and its impact on the TMI-2 program. The benefits of expanding the TMI-2 experience to general use in the nuclear industry are discussed

  11. Final Report Independent Verification Survey of the High Flux Beam Reactor, Building 802 Fan House Brookhaven National Laboratory Upton, New York

    Energy Technology Data Exchange (ETDEWEB)

    Harpeneau, Evan M. [Oak Ridge Institute for Science and Education, Oak Ridge, TN (United States). Independent Environmental Assessment and Verification Program

    2011-06-24

    On May 9, 2011, ORISE conducted verification survey activities including scans, sampling, and the collection of smears of the remaining soils and off-gas pipe associated with the 802 Fan House within the HFBR (High Flux Beam Reactor) Complex at BNL. ORISE is of the opinion, based on independent scan and sample results obtained during verification activities at the HFBR 802 Fan House, that the FSS (final status survey) unit meets the applicable site cleanup objectives established for as left radiological conditions.

  12. LETTER REPORT - INDEPENDENT VERIFICATION OF THE HIGH FLUX BEAM REACTOR DECOMMISSIONING PROJECT FAN HOUSE, BUILDING 704 BROOKHAVEN NATIONAL LABORATORY UPTON, NEW YORK

    International Nuclear Information System (INIS)

    Weaver, P.C.

    2010-01-01

    Oak Ridge Institute for Science and Education (ORISE) personnel visited the Brookhaven National Laboratory (BNL) on August 17 through August 23, 2010 to perform visual inspections and conduct independent measurement and sampling of the 'Outside Areas' at the High Flux Beam Reactor (HFBR) decommissioning project. During this visit, ORISE was also able to evaluate Fan House, Building 704 survey units (SUs) 4 and 5, which are part of the Underground Utilities portion of the HFBR decommissioning project. ORISE performed limited alpha plus beta scans of the remaining Fan House foundation lower walls and remaining pedestals while collecting static measurements. Scans were performed using gas proportional detectors coupled to ratemeter-scalers with audible output and encompassed an area of approximately 1 square meter around the static measurement location. Alpha plus beta scans ranged from 120 to 460 cpm. Twenty smears for gross alpha and beta activity and tritium were collected at judgmentally selected locations on the walls and pedestals of the Fan House foundation. Attention was given to joints, cracks, and penetrations when determining each sample location. Removable concentrations ranged from -0.43 to 1.73 dpm/100 cm2 for alpha and -3.64 to 7.80 dpm/100 cm2 for beta. Tritium results for smears ranged from -1.9 to 9.0 pCi/g. On the concrete pad, 100% of accessible area was scanned using a large area alpha plus beta gas proportional detector coupled to a ratemeter-scaler. Gross scan count rates ranged from 800 to 1500 cpm using the large area detector. Three concrete samples were collected from the pad primarily for tritium analysis. Tritium concentrations in concrete samples ranged from 53.3 to 127.5 pCi/g. Gamma spectroscopy results of radionuclide concentrations in concrete samples ranged from 0.02 to 0.11 pCi/g for Cs-137 and 0.19 to 0.22 pCi/g for Ra-226. High density scans for gamma radiation levels were performed in accessible areas in each SU, Fan House

  13. Nuclear reactor containment device

    International Nuclear Information System (INIS)

    Ichiki, Tadaharu.

    1980-01-01

    Purpose: To reduce the volume of a containment shell and decrease the size of a containment equipment for BWR type reactors by connecting the containment shell and a suppression pool with slanted vent tubes to thereby shorten the vent tubes. Constitution: A pressure vessel containing a reactor core is installed at the center of a building and a containment vessel for the nuclear reactor that contains the pressure vessel forms a cabin. To a building situated below the containment shell, is provided a suppression chamber in which cooling water is charged to form a suppression pool. The suppression pool is communicated with vent tubes that pass through the partition wall of the containment vessel. The vent tubes are slanted and their lower openings are immersed in coolants. Therefore, if accident is resulted and fluid at high temperature and high pressure is jetted from the pressure vessel, the jetting fluid is injected and condensated in the cooling water. (Moriyama, K.)

  14. Fusion Reactor Materials

    Energy Technology Data Exchange (ETDEWEB)

    Decreton, M

    2002-04-01

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

  15. Reactor Physics

    Energy Technology Data Exchange (ETDEWEB)

    Ait Abderrahim, A

    2001-04-01

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

  16. Reactor Physics

    International Nuclear Information System (INIS)

    Ait Abderrahim, A.

    2001-01-01

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

  17. Reactor Physics

    Energy Technology Data Exchange (ETDEWEB)

    Ait Abderrahim, A

    2002-04-01

    SCK-CEN's Reactor Physics and MYRRHA Department offers expertise in various areas of reactor physics, in particular in neutron and gamma calculations, reactor dosimetry, reactor operation and control, reactor code benchmarking and reactor safety calculations. This expertise is applied in the Department's own research projects in the VENUS critical facility, in the BR1 reactor and in the MYRRHA project (this project aims at designing a prototype Accelerator Driven System). Available expertise is also used in programmes external to the Department such as the reactor pressure steel vessel programme, the BR2 materials testing reactor dosimetry, and the preparation and interpretation of irradiation experiments by means of neutron and gamma calculations. The activities of the Fuzzy Logic and Intelligent Technologies in Nuclear Science programme cover several domains outside the department. Progress and achievements in these topical areas in 2001 are summarised.

  18. Reactor safeguards

    CERN Document Server

    Russell, Charles R

    1962-01-01

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

  19. Reactor operation

    CERN Document Server

    Shaw, J

    2013-01-01

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

  20. Nuclear research reactors activities in INVAP

    International Nuclear Information System (INIS)

    Ordonez, Juan Pablo

    2013-01-01

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

  1. Research reactors: design, safety requirements and applications

    International Nuclear Information System (INIS)

    Hassan, Abobaker Mohammed Rahmtalla

    2014-09-01

    There are two types of reactors: research reactors or power reactors. The difference between the research reactor and energy reactor is that the research reactor has working temperature and fuel less than the power reactor. The research reactors cooling uses light or heavy water and also research reactors need reflector of graphite or beryllium to reduce the loss of neutrons from the reactor core. Research reactors are used for research training as well as testing of materials and the production of radioisotopes for medical uses and for industrial application. The difference is also that the research reactor smaller in terms of capacity than that of power plant. Research reactors produce radioactive isotopes are not used for energy production, the power plant generates electrical energy. In the world there are more than 284 reactor research in 56 countries, operates as source of neutron for scientific research. Among the incidents related to nuclear reactors leak radiation partial reactor which took place in three mile island nuclear near pennsylvania in 1979, due to result of the loss of control of the fission reaction, which led to the explosion emitting hug amounts of radiation. However, there was control of radiation inside the building, and so no occurred then, another accident that lead to radiation leakage similar in nuclear power plant Chernobyl in Russia in 1986, has led to deaths of 4000 people and exposing hundreds of thousands to radiation, and can continue to be effect of harmful radiation to affect future generations. (author)

  2. LMFBR type reactor

    International Nuclear Information System (INIS)

    Kumaoka, Yoshio; Kawamura, Yutaka.

    1990-01-01

    A main vessel support skirt is supported by a base mat of reactor buildings and the base mat is supported by means of an earthquake-proof device on concretes of a lower raft disposed on ground rocks. The earthquake-proof device is constituted by alternately stacking, laminating and press-bonding thin steel plates together with thin rubber layers between an upper flange and a lower raft. Thus, for the horizontal seismic vibrations, the period of the swinging in the horizontal direction of the buildings is made greater than the swinging period of earthquakes by the earthquake-proof device to reduce the impact shocks of earthquakes. Further, for the vertical seismic vibrations, the input seismic movements are not amplified during transmission from the base mat of the buildings to the reactor structure by way of the support skirt of the main vessel, due to the shortened load transmission path and the seismic power design to the reactor structure can be moderated sufficiently. A safety LMFBR type reactor with reduced construction cost and improved reliability can be attained. (N.H.)

  3. AREVA's nuclear reactors portfolio

    International Nuclear Information System (INIS)

    Marincic, A.

    2009-01-01

    A reasonable assumption for the estimated new build market for the next 25 years is over 340 GWe net. The number of prospect countries is growing almost each day. To address this new build market, AREVA is developing a comprehensive portfolio of reactors intended to meet a wide range of power requirements and of technology choices. The EPR reactor is the flagship of the fleet. Intended for large power requirements, the four first EPRs are being built in Finland, France and China. Other countries and customers are in view, citing just two examples: the Usa where the U.S. EPR has been selected as the technology of choice by several U.S utilities; and the United Kingdom where the Generic Design Acceptance process of the EPR design submitted by AREVA and EDF is well under way, and where there is a strong will to have a plant on line in 2017. For medium power ranges, the AREVA portfolio includes a boiling water reactor and a pressurized water reactor which both offer all of the advantages of an advanced plant design, with excellent safety performance and competitive power generation cost: -) KERENA (1250+ MWe), developed in collaboration with several European utilities, and in particular with Eon; -) ATMEA 1 (1100+ MWe), a 3-loop evolutionary PWR which is being developed by AREVA and Mitsubishi. AREVA is also preparing the future and is deeply involved into Gen IV concepts. It has developed the ANTARES modular HTR reactor (pre-conceptual design completed) and is building upon its vast Sodium Fast Reactor experience to take part into the development of the next prototype. (author)

  4. Research Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Martens, Frederick H. [Argonne National Laboratory; Jacobson, Norman H.

    1968-09-01

    This booklet discusses research reactors - reactors designed to provide a source of neutrons and/or gamma radiation for research, or to aid in the investigation of the effects of radiation on any type of material.

  5. Nuclear reactors

    International Nuclear Information System (INIS)

    Middleton, J.E.

    1977-01-01

    Reference is made to water cooled reactors and in particular to the cooling system of steam generating heavy water reactors (SGHWR). A two-coolant circuit is described for the latter. Full constructural details are given. (U.K.)

  6. Reactor Neutrinos

    OpenAIRE

    Kim, Soo-Bong; Lasserre, Thierry; Wang, Yifang

    2013-01-01

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

  7. BOILING REACTORS

    Science.gov (United States)

    Untermyer, S.

    1962-04-10

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

  8. Reactor vessel

    NARCIS (Netherlands)

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

    1999-01-01

    A reactor vessel (1) comprises a reactor body (2) through which channels (3) are provided whose surface comprises longitudinal inwardly directed parts (4) and is provided with a catalyst (6), as well as buffer bodies (8, 12) connected to the channels (3) on both sides of the reactor body (2) and

  9. US Department of Energy Three Mile Island research and development program: Annual report, 1986

    Energy Technology Data Exchange (ETDEWEB)

    None

    1987-04-01

    Defueling of the Three Mile Island Unit 2 (TMI-2) reactor continued through 1986. This report summarizes this work and other TMI-2 related cleanup, research, and development activities. Other major topics include: core stratification sampling and other data acquisition tasks, the fuel shipping program, waste immobilization and management, decontamination and dose reduction, and future uses and applications of TMI-2 data.

  10. Nuclear power reactors

    International Nuclear Information System (INIS)

    1982-11-01

    After an introduction and general explanation of nuclear power the following reactor types are described: magnox thermal reactor; advanced gas-cooled reactor (AGR); pressurised water reactor (PWR); fast reactors (sodium cooled); boiling water reactor (BWR); CANDU thermal reactor; steam generating heavy water reactor (SGHWR); high temperature reactor (HTR); Leningrad (RMBK) type water-cooled graphite moderated reactor. (U.K.)

  11. Nuclear research reactors in Brazil

    Energy Technology Data Exchange (ETDEWEB)

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

    2011-07-01

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

  12. Nuclear research reactors in Brazil

    International Nuclear Information System (INIS)

    Cota, Anna Paula Leite; Mesquita, Amir Zacarias

    2011-01-01

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

  13. Nuclear Reactor RA Safety Report, Vol. 14, Safety protection measures

    International Nuclear Information System (INIS)

    1986-11-01

    Nuclear reactor accidents can be caused by three type of errors: failure of reactor components including (1) control and measuring instrumentation, (2) errors in operation procedure, (3) natural disasters. Safety during reactor operation are secured during its design and construction and later during operation. Both construction and administrative procedures are applied to attain safe operation. Technical safety features include fission product barriers, fuel elements cladding, primary reactor components (reactor vessel, primary cooling pipes, heat exchanger in the pump), reactor building. Safety system is the system for safe reactor shutdown and auxiliary safety system. RA reactor operating regulations and instructions are administrative acts applied to avoid possible human error caused accidents [sr

  14. Reactor physics and reactor computations

    International Nuclear Information System (INIS)

    Ronen, Y.; Elias, E.

    1994-01-01

    Mathematical methods and computer calculations for nuclear and thermonuclear reactor kinetics, reactor physics, neutron transport theory, core lattice parameters, waste treatment by transmutation, breeding, nuclear and thermonuclear fuels are the main interests of the conference

  15. Research reactors

    International Nuclear Information System (INIS)

    Merchie, Francois

    2015-10-01

    This article proposes an overview of research reactors, i.e. nuclear reactors of less than 100 MW. Generally, these reactors are used as neutron generators for basic research in matter sciences and for technological research as a support to power reactors. The author proposes an overview of the general design of research reactors in terms of core size, of number of fissions, of neutron flow, of neutron space distribution. He outlines that this design is a compromise between a compact enough core, a sufficient experiment volume, and high enough power densities without affecting neutron performance or its experimental use. The author evokes the safety framework (same regulations as for power reactors, more constraining measures after Fukushima, international bodies). He presents the main characteristics and operation of the two families which represent almost all research reactors; firstly, heavy water reactors (photos, drawings and figures illustrate different examples); and secondly light water moderated and cooled reactors with a distinction between open core pool reactors like Melusine and Triton, pool reactors with containment, experimental fast breeder reactors (Rapsodie, the Russian BOR 60, the Chinese CEFR). The author describes the main uses of research reactors: basic research, applied and technological research, safety tests, production of radio-isotopes for medicine and industry, analysis of elements present under the form of traces at very low concentrations, non destructive testing, doping of silicon mono-crystalline ingots. The author then discusses the relationship between research reactors and non proliferation, and finally evokes perspectives (decrease of the number of research reactors in the world, the Jules Horowitz project)

  16. Reactor container

    International Nuclear Information System (INIS)

    Ichiki, Tadaharu; Nagatomi, Shozo.

    1976-01-01

    Object: To provide a jet and missile protective wall of a configuration being inflated toward the center of a reactor container on the inside of a body of the reactor container disposed within a biological shield wall to thereby increase safety of the reactor container. Structure: A jet and missile protective wall comprised of curved surfaces internally formed with a plurality of arch inflations filled with concrete between inner and outer iron plates and shape steel beam is provided between a reactor container surrounded by a biological shield wall and a thermal shield wall surrounding the reactor pressure vessel, and an adiabatic heat insulating material is filled in space therebetween. (Yoshino, Y.)

  17. US Department of Energy Three Mile Island Research and Development Program: 1987 annual report

    International Nuclear Information System (INIS)

    1988-04-01

    Defueling of the Three Mile Island Unit 2 (TMI-2) reactor continued through 1987. This report summarizes this work and other TMI-2 related cleanup, research, and development activities. Other major topics include: Waste immobilization; Core transportation, receipt, and storage; Abnormal waste; Accident Evaluation and Technical Integration Programs; and Future uses and applications of TMI-2 data. While the technology being developed is of direct benefit to the recovery operations at TMI-2, it will also benefit the entire nuclear power industry

  18. Contribution to the prediction of sodium-water reactions effects: application to confinement losses inside a steam generator building of a sodium fast reactor

    International Nuclear Information System (INIS)

    Daudin, Kevin

    2015-01-01

    Study of sodium-water reaction (SWR) consequences in open air represents a challenge in the frame of safety assessments of sodium fast reactors (SFR). In case of major accident and to predict consequences of SWR, it is necessary to better appreciate phenomena and especially quantity and rate of the energy release. The objective is thus to strengthen the understanding of such reactions in order to predict with lore accuracy its consequences on mechanical equipment in the surroundings. This work focuses on three areas : research of accidental sequences, experimental investigation, and phenomenological analysis before the explosive contact. At first, a tree structure risk analysis with calculations of dangerous phenomena permitted to suggest how the contact between reactants may happen. Then, demonstrative experimental studies were performed to deepen some practical aspects of the phenomenology, like the influence of the way the reactants get in contact. Data analysis conducted to the development of a phenomenological model, implemented into a software platform for numerical simulations. Although numerous hypothesis, transient heat transfer consideration enables to reproduce experimental observations, especially the influence of mixing conditions (sodium mass and initial temperatures) on the phenomenology. This study of the premixing step of sodium-water explosion is relevant in the frame of current prediction methods of mechanical loadings on structures. (author) [fr

  19. Building founding method

    International Nuclear Information System (INIS)

    Tateno, Seiya; Hirako, Shizuka.

    1996-01-01

    The lower portion of a reactor building is buried at a level of a base rock. In the step of building construction, the base rock around the building is dug of keeping a construction space. The base rock present between the digging line and the building is replaced with an artificial base rock comprising aggregates and cement having a building constraining force substantially equal with that of the lower supporting base rock to obtain constraining force for the side portion of the building. The building itself is made of concrete steels. As the material for replacing the base rock between the digging line and the building, non-reinforced concrete having no steels may provide sufficient function. The burying depth is determined so as to align the surface of the base rock with the level of the base floor of the building. With such procedures, since it is not necessary to increase the area of the lower portion of the building, the amount of the materials for the building can be reduced. In addition, the earthquakes stability can be ensured without changing the shape of the building. (I.N.)

  20. Research reactors

    International Nuclear Information System (INIS)

    Kowarski, L.

    1955-01-01

    It brings together the techniques data which are involved in the discussion about the utility for a research institute to acquire an atomic reactor for research purposes. This type of decision are often taken by non-specialist people who can need a brief presentation of a research reactor and its possibilities in term of research before asking advises to experts. In a first part, it draws up a list of the different research programs which can be studied by getting a research reactor. First of all is the reactor behaviour and kinetics studies (reproducibility factor, exploration of neutron density, effect of reactor structure, effect of material irradiation...). Physical studies includes study of the behaviour of the control system, studies of neutron resonance phenomena and study of the fission process for example. Chemical studies involves the study of manipulation and control of hot material, characterisation of nuclear species produced in the reactor and chemical effects of irradiation on chemical properties and reactions. Biology and medicine research involves studies of irradiation on man and animals, genetics research, food or medical tools sterilization and neutron beams effect on tumour for example. A large number of other subjects can be studied in a reactor research as reactor construction material research, fabrication of radioactive sources for radiographic techniques or applied research as in agriculture or electronic. The second part discussed the technological considerations when choosing the reactor type. The technological factors, which are considered for its choice, are the power of the reactor, the nature of the fuel which is used, the type of moderator (water, heavy water, graphite or BeO) and the reflector, the type of coolants, the protection shield and the control systems. In the third part, it described the characteristics (place of installation, type of combustible and comments) and performance (power, neutron flux ) of already existing

  1. International Thermonuclear Experimental Reactor

    International Nuclear Information System (INIS)

    Blevins, J.D.; Stasko, R.R.

    1989-09-01

    An international design team comprised of members from Canada, Europe, Japan, the Soviet Union, and the United States of America, are designing an experimental fusion test reactor. The engineering and testing objectives of this International Thermonuclear Experimental Reactor (ITER) are to validate the design and to demonstrate controlled ignition, extended burn of a deuterium and tritium plasma, and achieve steady state using technology expected to be available by 1990. The concept maximizes flexibility while allowing for a variety of plasma configurations and operating scenarios. During physics phase operation, the machine produces a 22 MA plasma current. In the technology phase, the machine can be reconfigured with a thicker shield and a breeding blanket to operate with an 18 MA plasma current at a major radius of 5.5 meters. Canada's involvement in the areas of safety, facility design, reactor configuration and maintenance builds on our internationally recognized design and operational expertise in developing tritium processes and CANDU related technologies

  2. Nuclear Reactor RA Safety Report, Vol. 4, Reactor

    International Nuclear Information System (INIS)

    1986-11-01

    RA research reactor is thermal heavy water moderated and cooled reactor. Metal uranium 2% enriched fuel elements were used at the beginning of its operation. Since 1976, 80% enriched uranium oxide dispersed in aluminium fuel elements were gradually introduced into the core and are the only ones presently used. Reactor core is cylindrical, having diameter 40 cm and 123 cm high. Reaktor core is made up of 82 fuel elements in aluminium channels, lattice is square, lattice pitch 13 cm. Reactor vessel is cylindrical made of 8 mm thick aluminium, inside diameter 140 cm and 5.5 m high surrounded with neutron reflector and biological shield. There is no containment, the reactor building is playing the shielding role. Three pumps enable circulation of heavy water in the primary cooling circuit. Degradation of heavy water is prevented by helium cover gas. Control rods with cadmium regulate the reactor operation. There are eleven absorption rods, seven are used for long term reactivity compensation, two for automatic power regulation and two for safety shutdown. Total anti reactivity of the rods amounts to 24%. RA reactor is equipped with a number of experimental channels, 45 vertical (9 in the core), 34 in the graphite reflector and two in the water biological shield; and six horizontal channels regularly distributed in the core. This volume include detailed description of systems and components of the RA reactor, reactor core parameters, thermal hydraulics of the core, fuel elements, fuel elements handling equipment, fuel management, and experimental devices [sr

  3. An analysis of the estimated capital cost of a fusion reactor

    International Nuclear Information System (INIS)

    Hollis, A.A.

    1981-06-01

    The cost of building a fusion reactor similar to the Culham Conceptual Tokamak reactor Mark IIB is assessed and compared with other published capital costs of fusion and fission reactors. It is concluded that capital-investment and structure-renewal costs for a typical fusion reactor as presently conceived are likely to be higher than for thermal-fission reactors. (author)

  4. Hybrid reactors

    International Nuclear Information System (INIS)

    Moir, R.W.

    1980-01-01

    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 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 -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

  5. The breeder reactor and Europe

    International Nuclear Information System (INIS)

    Daglish, J.

    1979-01-01

    A report is given of a conference on the breeder reactor and Europe held in Lucerne, Switzerland from 14 - 17 October 1979 sponsored by the Swiss Association for Atomic Energy and the Association of European Atomic Forums. The underlying theme of the conference was the question that if nuclear power is to play a major role in meeting world energy needs in the long term, thermal reactors must in time be complemented with more advanced reactor systems that conserve uranium resources which are huge but not unlimited. This is not questioned; disagreement begins with discussion of the desirability of the breeder, and how fast and how far the introduction of such reactors should go. Aspects considered at the conference which are especially dealt with in this review are; why breed, commercial aspects, alternatives to the LMFBR, how to build a fast reactor, the breeder programmes in Europe, Britain, the Soviet Union, Japan and the United States. (U.K.)

  6. Jules Horowitz reactor (RJH): its design

    International Nuclear Information System (INIS)

    Dupuy, J.P.

    2002-01-01

    This article presents the design of the new irradiation facility (Jules Horowitz reactor) that is planned to be built on the Cadarache site of Cea. 2 principles have been followed. The first one is based on a physical separation between the systems and activities related to the reactor and the experiments from one hand and the other systems and means dedicated to the treatment of the experimental devices before and after irradiation on the other hand. This first principle implies to build 2 buildings: the reactor building and the nuclear auxiliaries building. Inside the reactor building activities from the reactor itself are separated from those dedicated to experimentation. In order to maximize the efficiency of such a reactor, an important number of simultaneous experiments is expected, which will generate an endless flux of incoming and out-going experiments and as a consequence an important handling work between the different work posts. The second principle aims at easing any handling work without breaking the rules of confinement. The different storing pools, the water pits that lead to the 5 hot cells and the reactor tank will communicate through a water-filled canal that will link the 2 buildings. (A.C.)

  7. Industrial structure at research reactor suppliers

    International Nuclear Information System (INIS)

    Roegler, H.-J.; Bogusch, E.; Friebe, T.

    2001-01-01

    Due to the recent joining of the forces of Framatome S. A. from France and the Nuclear Division of Siemens AG Power Generation (KWU) from Germany to a Joint Venture named Framatome Advanced Nuclear Power S.A.S., the issue of the necessary and of the optimal industrial structure for nuclear projects as a research reactor is, was discussed internally often and intensively. That discussion took place also in the other technical fields such as Services for NPPs but also in the field of interest here, i. e. Research Reactors. In summarizing the statements of this presentation one can about state that: Research Reactors are easier to build than NPPs, but not standardised; Research Reactors need a wide spectrum of skills and experiences; to design and build Research Reactors needs an experienced team especially in terms of management and interfaces; Research Reactors need background from built reference plants more than from operating plants; Research Reactors need knowledge of suitable experienced subsuppliers. Two more essential conclusions as industry involved in constructing and upgrading research reactors are: Research Reactors by far are more than a suitable core that generates a high neutron flux; every institution that designs and builds a Research Reactor lacks quality or causes safety problems, damages the reputation of the entire community

  8. Heterogeneous reactors

    International Nuclear Information System (INIS)

    Moura Neto, C. de; Nair, R.P.K.

    1979-08-01

    The microscopic study of a cell is meant for the determination of the infinite multiplication factor of the cell, which is given by the four factor formula: K(infinite) = n(epsilon)pf. The analysis of an homogeneous reactor is similar to that of an heterogeneous reactor, but each factor of the four factor formula can not be calculated by the formulas developed in the case of an homogeneous reactor. A great number of methods was developed for the calculation of heterogeneous reactors and some of them are discussed. (Author) [pt

  9. Report of the Federal Ministry for the Environment, Nature Conservation, Buildings and Nuclear Safety (BMUB) on the topical peer review aging management in nuclear power plants and research reactors

    International Nuclear Information System (INIS)

    2017-01-01

    The report of the Federal Environmental Ministry (BMUB) on the topical peer review aging management in nuclear power plants and research reactors covers the following issues: comprehensive requirements for aging management and its implementation, electric cables, non accessible pipes, reactor pressure vessel, calandria/pressure tubes (CANDU), concrete containment, pre-stressed concrete reactor pressure vessel (AGR).

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2014-10-15

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

  11. Nuclear reactor

    International Nuclear Information System (INIS)

    Mysels, K.J.; Shenoy, A.S.

    1976-01-01

    A nuclear reactor is described in which the core consists of a number of fuel regions through each of which regulated coolant flows. The coolant from neighbouring fuel regions is combined in a manner which results in an averaging of the coolant temperature at the outlet of the core. By this method the presence of hot streaks in the reactor is reduced. (UK)

  12. Nuclear reactor

    International Nuclear Information System (INIS)

    Tilliette, Z.

    1975-01-01

    A description is given of a nuclear reactor and especially a high-temperature reactor in which provision is made within a pressure vessel for a main cavity containing the reactor core and a series of vertical cylindrical pods arranged in spaced relation around the main cavity and each adapted to communicate with the cavity through two collector ducts or headers for the primary fluid which flows downwards through the reactor core. Each pod contains two superposed steam-generator and circulator sets disposed in substantially symmetrical relation on each side of the hot primary-fluid header which conveys the primary fluid from the reactor cavity to the pod, the circulators of both sets being mounted respectively at the bottom and top ends of the pod

  13. The fast breeder reactor Rapsodie (1962)

    International Nuclear Information System (INIS)

    Vautrey, L.; Zaleski, C.P.

    1962-01-01

    In this report, the authors describe the Rapsodie project, the French fast breeder reactor, as it stands at construction actual start-up. The paper provides informations about: the principal neutronic and thermal characteristics, the reactor and its cooling circuits, the main handling devices of radioactive or contaminated assemblies, the principles and means governing reactor operation, the purposes and locations of miscellaneous buildings. Rapsodie is expected to be critical by 1964. (authors) [fr

  14. Exporting apocalypse: CANDU reactors and nuclear proliferation

    International Nuclear Information System (INIS)

    McKay, Paul.

    The author believes that the peaceful use of nuclear technology leads inevitably to the production of nuclear weapons, and that CANDU reactors are being bought by countries that are likely to build bombs. He states that exports of reactors and nuclear materials cannot be defended and must be stopped

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2013-10-15

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

  16. The reactor Cabri

    International Nuclear Information System (INIS)

    Ailloud, J.; Millot, J.P.

    1964-01-01

    It has become necessary to construct in France a reactor which would permit the investigation of the conditions of functioning of future installations, the choice, the testing and the development of safety devices to be adopted. A water reactor of a type corresponding to the latest CEA constructions in the field of laboratory or university reactors was decided upon: it appeared important to be able to evaluate the risks entailed and to study the possibilities of increasing the power, always demanded by the users; on the other hand, it is particularly interesting to clarify the phenomena of power oscillation and the risks of burn out. The work programme for CABRI will be associated with the work carried out on the American Sperts of the same type, during its construction, very useful contacts were made with the American specialists who designed the se reactors. A brief description of the reactor is given in the communication as well as the work programme for the first years with respect to the objectives up to now envisaged. Rough description of the reactor. CABRI is an open core swimming-pool reactor without any lateral protection, housed in a reinforced building with controlled leakage, in the Centre d'Etudes Nucleaires de Cadarache. It lies alone in the middle of an area whose radius is 300 meters long. Control and measurements equipment stand out on the edge of that zone. It consumes MTR fuel elements. The control-safety rods are propelled by compressed air. The maximum flow rate of cooling circuit is 1500 m 3 /h. Transient measurements are recorded in a RW330 unit. Aims and work programme. CABRI is meant for: - studies on the safety of water reactors - for the definition of the safety margins under working conditions: research of maximum power at which a swimming-pool reactor may operate with respect to a cooling accident, of local boiling effect on the nuclear behaviour of the reactor, performances of the control and safety instruments under exceptional

  17. SLOWPOKE reactor

    International Nuclear Information System (INIS)

    Evans, D.J.R.; Downs, W.E.

    1974-01-01

    The SLOWPOKE reactor is described, which is a small pool type with thermal neutron fluxes ranging from 10 11 -10 12 n cm -2 sec -1 . It differs in many ways from conventional pool type, namely small critical mass, beryllium reflector and a closed reactor container. The reactor is designed as small and simply as possible, and consistently with safety and good operating practice. Access to the present model is via pneumatic irradiation tubes only, which limits the use of the facility to activation analysis, tracer production and training. (Mori, K.)

  18. Nuclear reactor

    International Nuclear Information System (INIS)

    Rau, P.

    1980-01-01

    The reactor core of nuclear reactors usually is composed of individual elongated fuel elements that may be vertically arranged and through which coolant flows in axial direction, preferably from bottom to top. With their lower end the fuel elements gear in an opening of a lower support grid forming part of the core structure. According to the invention a locking is provided there, part of which is a control element that is movable along the fuel element axis. The corresponding locking element is engaged behind a lateral projection in the opening of the support grid. The invention is particularly suitable for breeder or converter reactors. (orig.) [de

  19. Safety of research reactors (Design and Operation)

    International Nuclear Information System (INIS)

    Dirar, H. M.

    2012-06-01

    The primary objective of this thesis is to conduct a comprehensive up-to-date literature review on the current status of safety of research reactor both in design and operation providing the future trends in safety of research reactors. Data and technical information of variety selected historical research reactors were thoroughly reviewed and evaluated, furthermore illustrations of the material of fuel, control rods, shielding, moderators and coolants used were discussed. Insight study of some historical research reactors was carried with considering sample cases such as Chicago Pile-1, F-1 reactor, Chalk River Laboratories,. The National Research Experimental Reactor and others. The current status of research reactors and their geographical distribution, reactor category and utilization is also covered. Examples of some recent advanced reactors were studied like safety barriers of HANARO of Korea including safety doors of the hall and building entrance and finger print identification which prevent the reactor from sabotage. On the basis of the results of this research, it is apparent that a high quality of safety of nuclear reactors can be attained by achieving enough robust construction, designing components of high levels of efficiency, replacing the compounds of the reactor in order to avoid corrosion and degradation with age, coupled with experienced scientists and technical staffs to operate nuclear research facilities.(Author)

  20. Reactor operational transient analysis

    International Nuclear Information System (INIS)

    Shin, W.K.; Chae, S.K.; Han, K.I.; Yang, K.S.; Chung, H. D.; Kim, H.G.; Moon, H.J.; Ryu, Y.H.

    1983-01-01

    To build up efficient capability of safety review and inspection for the nuclear power plants, four area of studies have performed as follows: 1) In order to search the most optimized operating method during load follow operating schemes, automatic control and normal control, are compared each other under the CAOC condition. The analysis performed by DDID code has shown that the reactor has to be controlled by the operator manually during load follow operation. 2) Through the sensitivity analysis by COBRA code, the operating parameters, such as coolant pressure, flow rate, inlet temperature, and power distribution are shown to be important to the determination of DNBR. Expecially, inlet temperature of primary coolant system is appeared as the most senstive parameter on DNBR. 3) FRAPCON code is adapted to study the sensitivity of several operational parameters on the mechanical properties of reactor fuel rod. 4) The calculations procedure which is required to be obtained the neutron fluence at the reactor vessel and the spectrum at the surveillance capsule is established. The results of computation are conpared with those of FSAR and SWRI report and proved its applicability to reactor surveillance program. (Author)

  1. NUCLEAR REACTOR

    Science.gov (United States)

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

    1958-01-21

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

  2. Reactor Neutrinos

    Directory of Open Access Journals (Sweden)

    Soo-Bong Kim

    2013-01-01

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

  3. REACTOR SHIELD

    Science.gov (United States)

    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.

  4. NEUTRONIC REACTOR

    Science.gov (United States)

    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.

  5. Chemical Reactors.

    Science.gov (United States)

    Kenney, C. N.

    1980-01-01

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

  6. Development of post-irradiation examination and evaluation techniques for nuclear reactor fuel (III)

    International Nuclear Information System (INIS)

    Ro, Seung Gy; Min, Duck Kee; Kim, Eun Ka; Lee, Young Gil; You, Gil Sung; Park, Yun Gyoo; Chun, Yong Bum; Koo, Dae Seo; Sohn, Dong Seong; Song, Keun Woo; Joe, Kih Soo

    1992-07-01

    In-pool and hot cell examinations have been carried out for the J14 fuel assembly and three fuel rods of G23 fuel assembly from the Kori-1 NPP at the KAERI's PIE facility. Numerous of PIE data on the high burnup fuels have been obtained. In-core performance of Kori-1 fuels has been evaluated based on the PIE data of 15 fuel rods irradiated for 1 to 4 reactor cycles in the Kori-1 NPP. The major in-core performance parameters considered were a clad creep deformation, a rod growth, a fission gas release, densification and swelling, and a clad corrosion. As a part of the collaboration in work on PIE between Korea and USA, hot cell examinations and evaluation have been carried out for five of 13 TMI-2 core debris samples received from INEL, USA. Burnup measurements of Kori-1 fuels have been performed with the high performance liquid chromatographic(HPLC) separation system. It was found that the measured burnups by the HPLC method are in good agreement with that obtained by the anion exchange method within ±2%. The reassembling device of dismantled fuel assembly has been designed and fabricated. The hole-down spring force measuring device has been modified to improve the accuracy of measuring system by replacing the hydraulic controlled-system with motor driven-system. (Author)

  7. Method for temporary shielding of reactor vessel internals

    International Nuclear Information System (INIS)

    Grimm, N.P.; Sejvar, J.

    1991-01-01

    This patent describes a method for shielding stored internals for reactor vessel annealing. It comprises removing nuclear fuel from the reactor vessel containment building; removing and storing upper and lower core internals under water in a refueling canal storage area; assembling a support structure in the refueling canal between the reactor vessel and the stored internals; introducing vertical shielding tanks individually through a hatch in the containment building and positioning each into the support structure; introducing horizontal shielding tanks individually through a hatch in the containment building and positioning each above the stored internals and vertical tanks; draining water from the refueling canal to the level of a flange of the reactor vessel; placing an annealing apparatus in the reactor vessel; pumping the remaining water from the reactor vessel; and annealing the reactor vessel

  8. Reactor technology

    International Nuclear Information System (INIS)

    Erdoes, P.

    1977-01-01

    This is one of a series of articles discussing aspects of nuclear engineering ranging from a survey of various reactor types for static and mobile use to mention of atomic thermo-electric batteries of atomic thermo-electric batteries for cardiac pacemakers. Various statistics are presented on power generation in Europe and U.S.A. and economics are discussed in some detail. Molten salt reactors and research machines are also described. (G.M.E.)

  9. N Reactor Deactivation Program Plan

    International Nuclear Information System (INIS)

    Walsh, J.L.

    1993-12-01

    This N Reactor Deactivation Program Plan is structured to provide the basic methodology required to place N Reactor and supporting facilities · in a radiologically and environmentally safe condition such that they can be decommissioned at a later date. Deactivation will be in accordance with facility transfer criteria specified in Department of Energy (DOE) and Westinghouse Hanford Company (WHC) guidance. Transition activities primarily involve shutdown and isolation of operational systems and buildings, radiological/hazardous waste cleanup, N Fuel Basin stabilization and environmental stabilization of the facilities. The N Reactor Deactivation Program covers the period FY 1992 through FY 1997. The directive to cease N Reactor preservation and prepare for decommissioning was issued by DOE to WHC on September 20, 1991. The work year and budget data supporting the Work Breakdown Structure in this document are found in the Activity Data Sheets (ADS) and the Environmental Restoration Program Baseline, that are prepared annually

  10. Economic analysis of EBT reactor

    International Nuclear Information System (INIS)

    Woo, J.T.; Uckan, N.A.; Lidsky, L.M.

    1977-01-01

    In order to establish the economic potential of the Elmo Bumpy Torus (EBT) reactor, two independent system-costing models have been developed. Both models predict capital costs of approximately $400/kW(th). These relatively low costs reflect the simplicity of the EBTR design. In particular, the modular nature of the individual blanket-shield segments, the low costs ''accelerator style'' containment building, high beta, and steady-state operation lead to relatively low reactor costs. A detailed cost breakdown for subsystems is analyzed. High cost and high uncertainty subsystems are identified to direct further emphasis into those areas. The calculated capital costs for the EBT reactor are compared with those costs quoted for tokamak reactors

  11. Propulsion reactors

    International Nuclear Information System (INIS)

    Anon.

    1999-01-01

    A nuclear reactor equips the recently constructed French aircraft- carrier Charles-De-Gaulle, in a few months the second nuclear submarine (SNLE) of new generation will be operational. In last october the government launched the program Barracuda which consists of 6 submarines (SNA) whose series head will be operational in 2010. The main asset of nuclear propulsion is to allow an almost unlimited autonomy: soft water, air are produced inside the submarine and the maximum time spent underwater is only limited by human capacity to cope with confinement. CEA has 3 missions concerning country defence. First the designing, the fabrication and the maintenance of weapons, secondly the supplying of fissile materials and thirdly the nuclear propulsion. A new generation of propulsion reactors is being studied and a ground installation involving a test reactor equivalent to that on board is being built. This test reactor (RES) will simulate any type of on-board reactors by adjusting temperature, pressure, flowrate and even equipment such as steam generator. This reactor will validate the technological choices for the Barracuda program. (A.C.)

  12. Advanced boiling water reactor

    International Nuclear Information System (INIS)

    Nishimura, N.; Nakai, H.; Ross, M.A.

    1999-01-01

    In the Boiling Water Reactor (BWR) system, steam generated within the nuclear boiler is sent directly to the main turbine. This direct cycle steam delivery system enables the BWR to have a compact power generation building design. Another feature of the BWR is the inherent safety that results from the negative reactivity coefficient of the steam void in the core. Based on the significant construction and operation experience accumulated on the BWR throughout the world, the ABWR was developed to further improve the BWR characteristics and to achieve higher performance goals. The ABWR adopted 'First of a Kind' type technologies to achieve the desired performance improvements. The Reactor Internal Pump (RIP), Fine Motion Control Rod Drive (FMCRD), Reinforced Concrete Containment Vessel (RCCV), three full divisions of Emergency Core Cooling System (ECCS), integrated digital Instrumentation and Control (I and C), and a high thermal efficiency main steam turbine system were developed and introduced into the ABWR. (author)

  13. Compact fusion reactors

    CERN Multimedia

    CERN. Geneva

    2015-01-01

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

  14. RA Research reactor, Part 1, Operation and maintenance of the RA nuclear reactor for 1986

    International Nuclear Information System (INIS)

    Sotic, O.; Martinc, R.; Cupac, S.; Sulem, B.; Badrljica, R.; Majstorovic, D.; Sanovic, V.

    1986-01-01

    In order to enable future reliable operation of the RA reactor, according to new licensing regulations, three major tasks started in 1984 were fulfilled: building of the new emergency system, reconstruction of the existing ventilation system, and reconstruction of the power supply system. Simultaneously in 1985/1986 renewal of the instrumentation and reconstruction of the system for handling and storage of the spent fuel in the reactor building have started. Design projects for these tasks are almost finished and the reconstruction of both systems is expected to be finished until 1988 and mid 1989 respectively. RA reactor Safety report was finished according to the recommendations of the IAEA. Investments in 1986 were used for 8000 kg of heavy water, maintenance of reactor systems and supply of new components, reconstruction of reactor systems. This report includes 8 annexes concerning reactor operation, activities of services and financial issues [sr

  15. SOLASE: a conceptual laser fusion reactor design

    International Nuclear Information System (INIS)

    Conn, R.W.; Abdel-Khalik, S.I.; Moses, G.A.

    1977-12-01

    The SOLASE conceptual laser fusion reactor has been designed to elucidate the technological problems posed by inertial confinement fusion reactors. This report contains a detailed description of all aspects of the study including the physics of pellet implosion and burn, optics and target illumination, last mirror design, laser system analysis, cavity design, pellet fabrication and delivery, vacuum system requirements, blanket design, thermal hydraulics, tritium analysis, neutronics calculations, radiation effects, stress analysis, shield design, reactor and plant building layout, maintenance procedures, and power cycle design. The reactor is designed as a 1000 MW/sub e/ unit for central station electric power generation

  16. Research reactors - an overview

    International Nuclear Information System (INIS)

    West, C.D.

    1997-01-01

    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

  17. Nuclear reactors

    International Nuclear Information System (INIS)

    Prescott, R.F.; George, B.V.; Baglin, C.J.

    1978-01-01

    Reference is made to thermal insulation on the inner surfaces of containment vessels of fluid cooled nuclear reactors and particularly in situations where the thermal insulation must also serve a structural function and transmit substantial load forces to the surface which it covers. An arrangement is described that meets this requirement and also provides for core support means that favourably influences the flow of hot coolant from the lower end of the core into a plenum space in the hearth of the reactor. The arrangement comprises a course of thermally insulating bricks arranged as a mosaic covering a wall of the reactor and a course of thermally insulating tiles arranged as a mosaic covering the course of bricks. Full constructional details are given. (UK)

  18. Reactor utilization

    International Nuclear Information System (INIS)

    Zecevic, V.

    1963-01-01

    In 1962, the RA reactor was operated almost three times more than in 1961, producing total of 25 555 MWh. Diagram containing comparative data about reactor operation for 1960, 1961, and 1962, percent of fuel used and U-235 burnup shows increase in reactor operation. Number of samples irradiated was 659, number of experiments done was 16. mean powered level was 5.93 MW. Fuel was added into the core twice during the reporting year. In fact the core was increased from 56 to 68 fuel channels and later to 84 fuel channels. Fuel was added to the core when the reactivity worth decreased to the minimum operation level due to burnup. In addition to this 5 central fuel channels were exchanged with fresh fuel in february for the purpose of irradiation in the VISA-2 channel

  19. Bioconversion reactor

    Science.gov (United States)

    McCarty, Perry L.; Bachmann, Andre

    1992-01-01

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

  20. Thermonuclear reactor

    International Nuclear Information System (INIS)

    Araki, Takao; Saito, Yasushi.

    1987-01-01

    Purpose: To reduce the seismic wave responsivity of an exhaust duct shields thereby preventing the release of tritium in an evacuating device due to failures upon earthquakes. Constitution: The ends on the cutting side of upper outer exhaust duct shields of a thermonuclear reactor are connected with a plurality of support beams. In a case where seismic vibrations are exerted to such a thermonuclear reactor, since the ends on the cutting side are coupled with the support beams, vibrations of the upper outer exhaust duct shields are greatly restricted. Thus, since there is no more such a possibility, for example, that an exhaust duct connected to the upper portion of a reactor main body is greatly distorted due to the seismic response of the upper outside exhaust duct shields to result in the failure of the connection portion with a vacuum pump, the release of tritium due to failure of the evacuating device can be prevented. (Yoshino, Y.)

  1. Nuclear reactors design study and parameters calculation

    International Nuclear Information System (INIS)

    Morcos, H.N.

    2002-01-01

    the nuclear design a reactor core needs to determine a set of system parameters which will lead to safe, reliable and economical reactor operation at the rated power level over the desired core lifetime. the principal tools used in this task consist of a number of models of neutron behavior in the reactor that are implemented by a multiplicity of computer programs or codes used to simulate the nuclear behavior of the reactor core. the study of the interaction of the core power distributions with the time-dependent production or depletion of nuclei in the core is known as depletion or burn up analysis the main objective of the present thesis is to study the fuel depletion analysis under different reactor operating regimes and their influence on the build up of actinides and fission products (F P). therefore, one can estimate the optimum reactor-operating regime at which the accumulation of certain actinide isotope can reach maximum

  2. Overview of the Dalat Nuclear Research Reactor

    International Nuclear Information System (INIS)

    Nguyen Nhi Dien; Nguyen Thai Sinh; Luong Ba Vien

    2016-01-01

    The present reactor called Dalat Nuclear Research Reactor (DNRR) has been reconstructed from the former TRIGA Mark II reactor which was designed by General Atomic (GA, San Diego, California, USA), started building in early 1960s, put into operation in 1963 and operated until 1968 at nominal power of 250 kW. In 1975, all fuel elements of the reactor were unloaded and shipped back to the USA. The DNRR is a 500-kW pool-type research reactor using light water as both moderator and coolant. The reactor is used as a neutron source for the purposes of: (1) radioactive isotope production; (2) neutron activation analysis; and (3) research and training

  3. Reactor primary containment vessel

    International Nuclear Information System (INIS)

    Kodama, Tasuku

    1986-01-01

    Purpose: To insure safety by preventing the radioactivity release to outside environments in the event of an unforeseen accident when the gas temperature in the reactor container is high. Constitution: In the event of any unforeseen accident, a high-temperature gas in the reactor container will be discharged from a discharge port into the pool water through a suction piping system and a discharge piping system, and after being cooled with the pool water, the gas will be discharged into the reactor building. Thereafter, the cooled gas will be drawn into an emergency gas processing device through the suction piping system and removed of radioactivity, finally being discharged out into the external environment. Soluble radioactive maters will be absorbed and removed into the pool water while they are in contact with the gas in the pool water. This, therefore, can contribute much more toward the control of discharge of the radioactivity. Where it is required to improve solubility, a methyl iodine solution can be used as the pool water. A fuel storage pool may be used as the pool. (Kamimura, M.)

  4. Nuclear reactor

    International Nuclear Information System (INIS)

    Gilroy, J.E.

    1980-01-01

    An improved cover structure for liquid metal cooled fast breeder type reactors is described which it is claimed reduces the temperature differential across the intermediate grid plate of the core cover structure and thereby reduces its subjection to thermal stresses. (UK)

  5. Neutronic reactor

    International Nuclear Information System (INIS)

    Wende, C.W.J.

    1976-01-01

    The method of operating a water-cooled neutronic reactor having a graphite moderator is described which comprises flowing a gaseous mixture of carbon dioxide and helium, in which the helium comprises 40--60 volume percent of the mixture, in contact with the graphite moderator. 2 claims, 4 figures

  6. Neutronic reactor

    International Nuclear Information System (INIS)

    Wende, C.W.J.

    1976-01-01

    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

  7. Reactor licensing

    International Nuclear Information System (INIS)

    Harvie, J.D.

    2002-01-01

    This presentation discusses reactor licensing and includes the legislative basis for licensing, other relevant legislation , the purpose of the Nuclear Safety and Control Act, important regulations, regulatory document, policies, and standards. It also discusses the role of the CNSC, its mandate and safety philosophy

  8. Neutronic reactor

    International Nuclear Information System (INIS)

    Carleton, J.T.

    1977-01-01

    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. 3 claims, 6 figures

  9. Vibration-proof FBR type reactor

    International Nuclear Information System (INIS)

    Kawamura, Yutaka.

    1992-01-01

    In a reactor container in an FBR type reactor, an outer building and upper and lower portions of a reactor container are connected by a load transmission device made of a laminated material of rubber and steel plates. Each of the reactor container and the outer building is disposed on a lower raft disposed on a rock by way of a vibration-proof device made of a laminated material of rubber and steel plates. Vibration-proof elements for providing vertical eigen frequency of the vibration-proof system comprising the reactor building and the vibration-proof device within a range of 3Hz to 5Hz are used. That is, the peak of designed acceleration for response spectrum in the horizontal direction of the reactor structural portions is shifted to side of shorter period from the main frequency region of the reactor structure. Alternatively, rigidity of the vibration-proof elements is decreased to shift the peak to the side of long period from the main frequency region. Designed seismic force can be greatly reduced both horizontally and vertically, to reduce the wall thickness of the structural members, improve the plant economy and to ensure the safety against earthquakes. (N.H.)

  10. Building-wake diffusion

    International Nuclear Information System (INIS)

    Ramsdell, J.V.

    1988-01-01

    A recent review of building-wake diffusion models compares model predictions of centerline concentrations in building wakes with concentrations observed in experiments at seven nuclear reactors. In general, the model predictions are conservative in that they tend to predict concentrations that are greater than those actually observed. However, the models show little skill in accounting for variations in the observed concentrations. Analysis of experimental data indicates that the general form of the standard wake diffusion models is inconsistent with observed variations of concentrations in the wakes. The inconsistency is especially marked for ground-level releases. As an interim measure, multiple linear regression techniques have been used to develop a statistical building-wake model alternative to the current models. This paper describes the statistical wake model and compares it with other models

  11. Three Mile Island unit 2 vessel investigation project. Conclusions and significance

    International Nuclear Information System (INIS)

    Beckjord, E.S.

    1994-01-01

    At the conclusion of the TMI-2 Vessel Investigation Project, additional insights about the accident have been gained, specifically in the area of reactor vessel integrity and the conditions of the lower head of the reactor vessel. This paper discusses three topics: the evolving views about the TMI-2 accident scenario over time, the technical conclusions of the TMI-2 VIP (recovery of samples from the vessel lower head), and the broad significance of these findings (accident management). 4 refs

  12. Dismantling method for reactor pressure vessel and system therefor

    International Nuclear Information System (INIS)

    Hayashi, Makoto; Enomoto, Kunio; Kurosawa, Koichi; Saito, Hideyo.

    1994-01-01

    Upon dismantling of a reactor pressure vessel, a containment building made of concretes is disposed underground and a spent pressure vessel is contained therein, and incore structures are contained in the spent pressure vessel. Further, a plasma-welder and a pressing machine are disposed to a pool for provisionally placing reactor equipments in the reactor building for devoluming the incore structures by welding and compression. An overhead-running crane and rails therefor are disposed on the roof and the outer side of the reactor building for transporting the pressure vessel from the reactor building to the containment building. They may be contained in the containment building after incorporation of the incore structures into the pressure vessel at the outside of the reactor building. For the devoluming treatment, a combination of cutting, welding, pressing and the like are optically conducted. A nuclear power plant can be installed by using a newly manufactured nuclear reactor, with no requirement for a new site and it is unnecessary to provide a new radioactive waste containing facility. (N.H.)

  13. Reactor core of nuclear reactor

    International Nuclear Information System (INIS)

    Sasagawa, Masaru; Masuda, Hiroyuki; Mogi, Toshihiko; Kanazawa, Nobuhiro.

    1994-01-01

    In a reactor core, a fuel inventory at an outer peripheral region is made smaller than that at a central region. Fuel assemblies comprising a small number of large-diameter fuel rods are used at the central region and fuel assemblies comprising a great number of smalldiameter fuel rods are used at the outer peripheral region. Since a burning degradation rate of the fuels at the outer peripheral region can be increased, the burning degradation rate at the infinite multiplication factor of fuels at the outer region can substantially be made identical with that of the fuels in the inner region. As a result, the power distribution in the direction of the reactor core can be flattened throughout the entire period of the burning cycle. Further, it is also possible to make the degradation rate of fuels at the outer region substantially identical with that of fuels at the inner side. A power peak formed at the outer circumferential portion of the reactor core of advanced burning can be lowered to improve the fuel integrity, and also improve the reactor safety and operation efficiency. (N.H.)

  14. Diffusion in building wakes

    International Nuclear Information System (INIS)

    Ramsdell, J.V.

    1988-03-01

    Straight-line Gaussian models adequately describe atmospheric diffusion for many applications. They have been modified for use in estimating diffusion in building wakes by adding terms that include projected building area and by redefining the diffusion coefficients so that the coefficients have minimum values that are related to building dimensions. In a recent study, Ramsdell reviewed the building-wake dispersion models used by the Nuclear Regulatory Commission (NRC) in its control room habitability assessments. The review included comparison of model estimates of centerline concentrations with concentrations observed in experiments at seven nuclear reactors. In general, the models are conservative in that they tend to predict concentrations that are greater than those actually observed. However, the models show little skill in accounting for variations in the observed concentrations. Subsequently, the experimental data and multiples linear regression techniques have been used to develop a new building wake diffusion model. This paper describes the new building wake model and compares it with other models. 8 refs., 2 figs

  15. Research reactor education and training

    International Nuclear Information System (INIS)

    Gless, B.; Chanteux, P.

    2003-01-01

    CORYS T.E.S.S. and TECHNICATOME present in this document some of the questions that can be rightfully raised concerning education and training of nuclear facilities' staffs. At first, some answers illustrate the tackled generic topics: importance of training, building of a training program, usable tools for training purposes. Afterwards, this paper deals more specifically with research reactors as an actual training tool. The pedagogical advantages they can bring are illustrated through an example consisting in the description of the AZUR facility training capabilities followed by the detailed experiences CORYS T.E.S.S. and TECHNICATOME have both gathered and keeps on gaining using research reactors for training means. The experience shows that this incomparable training material is not necessarily reserved to huge companies or organisations' numerous personnel. It offers enough flexibility to be adapted to the specific needs of a thinner audience. Thus research reactor staffs can also take advantages of this training method. (author)

  16. Nuclear reactor

    International Nuclear Information System (INIS)

    Gibbons, J.F.; McLaughlin, D.J.

    1978-01-01

    In the pressure vessel of the water-cooled nuclear reactor there is provided an internal flange on which the one- or two-part core barrel is hanging by means of an external flange. A cylinder is extending from the reactor vessel closure downwards to a seat on the core cupport structure and serves as compression element for the transmission of the clamping load from the closure head to the core barrel (upper guide structure). With the core barrel, subject to tensile stress, between the vessel internal flange and its seat on one hand and the compression of the cylinder resp. hold-down element between the closure head and the seat on the other a very strong, elastic sprung structure is obtained. (DG) [de

  17. Nuclear reactor

    International Nuclear Information System (INIS)

    Aleite, W.; Bock, H.W.; Struensee, S.

    1976-01-01

    The invention concerns the use of burnable poisons in a nuclear reactor, especially in PWRs, in order to improve the controllability of the reactor. An unsymmetrical arrangement in the lattice is provided, if necessary also by insertion of special rods for these additions. It is proposed to arrange the burnable poisons in fuel elements taken over from a previous burn-up cycle and to distribute them, going out from the side facing the control rods, over not more than 20% of the lenth of the fuel elements. It seems sufficient, for the burnable poisons to bind an initial reactivity of only 0.1% and to become ineffective after normal operation of 3 to 4 months. (ORU) [de

  18. Reactor container

    International Nuclear Information System (INIS)

    Ichiki, Tadaharu; Saba, Kazuhisa.

    1979-01-01

    Purpose: To improve the earthquake resistance as well as reduce the size of a container for a nuclear reactor with no adverse effects on the decrease of impact shock to the container and shortening of construction step. Constitution: Reinforcing profile steel materials are welded longitudinally and transversely to the inner surface of a container, and inner steel plates are secured to the above profile steel materials while keeping a gap between the materials and the container. Reactor shielding wall planted to the base concrete of the container is mounted to the pressure vessel, and main steam pipeways secured by the transverse beams and led to the outside of container is connected. This can improve the rigidity earthquake strength and the safetiness against the increase in the inside pressure upon failures of the container. (Yoshino, Y.)

  19. Reactor container

    International Nuclear Information System (INIS)

    Oyamada, Osamu; Furukawa, Hideyasu; Uozumi, Hiroto.

    1979-01-01

    Purpose: To lower the position of an intermediate slab within a reactor container and fitting a heat insulating material to the inner wall of said intermediate slab, whereby a space for a control rod exchanging device and thermal stresses of the inner peripheral wall are lowered. Constitution: In the pedestal at the lower part of a reactor pressure vessel there is formed an intermediate slab at a position lower than diaphragm floor slab of the outer periphery of the pedestal thereby to secure a space for providing automatic exchanging device of a control rod driving device. Futhermore, a heat insulating material is fitted to the inner peripheral wall at the upper side of the intermediate slab part, and the temperature gradient in the wall thickness direction at the time of a piping rupture trouble is made gentle, and thermal stresses at the inner peripheral wall are lowered. (Sekiya, K.)

  20. Neutronic reactor

    International Nuclear Information System (INIS)

    Lewis, W.R.

    1978-01-01

    Disclosed is 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

  1. Nuclear reactor

    International Nuclear Information System (INIS)

    Shirakawa, Toshihisa.

    1979-01-01

    Purpose: To prevent cladding tube injuries due to thermal expansion of each of the pellets by successively extracting each of the control rods loaded in the reactor core from those having less number of notches, as well as facilitate the handling work for the control rods. Constitution: A recycle flow control device is provided to a circulation pump for forcibly circulating coolants in the reactor container and an operational device is provided for receiving each of the signals concerning number of notches for each of the control rods and flow control depending on the xenon poisoning effect obtained from the signals derived from the in-core instrument system connected to the reactor core. The operational device is connected with a control rod drive for moving each of the control rods up and down and a recycle flow control device. The operational device is set with a pattern for the aimed control rod power and the sequence of extraction. Upon extraction of the control rods, they are extracted successively from those having less notch numbers. (Moriyama, K.)

  2. US DOE Idaho national laboratory reactor decommissioning

    International Nuclear Information System (INIS)

    Szilagyi, Andrew

    2012-01-01

    , asbestos and mercury among others. Each reactor required isolation in order to be removed. Due to activated metal within the reactor vessels, dose rates above the cores ranged from 50 R/hr to 1200 R/hr. Subsequent dose rates outside the vessels varied from 60 mR/hr to greater than 50 R/hr. Due to the elevated dose rates, the project team decided to fill the ETR and MTR reactor vessels with grout to a level above the core region to reduce dose. To remove the ETR reactor, access to the support shoes was required. These shoes were encased in the high density concrete biological shield approximately 8' below grade. The project team used explosives to remove the biological shield. The demolition had to be controlled to prevent damaging the reactor vessel and to limit the seismic impact on a nearby operating reactor. Upon completion of the blast, the concrete was removed exposing the support shoes for the vessel. Two reactor buildings (ETR and PBF) had to be removed to accommodate lifting systems for the reactor vessels. Two reactors (PBF and MTR) were removed via mobile cranes, two reactors were sized and removed in pieces (ZPPR and MTR), and ETR reactor, due to its weight, was removed via a twin gantry lifting system

  3. Benefits of nuclear reactor still unclear

    International Nuclear Information System (INIS)

    Allen, Barry

    1997-01-01

    The author questions the Australian Government decision to build a new reactor at Lucas Heights and to reject the proposal for a nuclear waste reprocessing and disposal using Australia's Synroc technology. He argued that Australia should have looked to the future(Synroc) instead of investing in dated technology (Reactor) and sees Synroc technology having much more potential to generate foreign currency if the increasing need for waste disposal facilities in the region are considered

  4. Effects of base mat flexibility and structure-soil-structure interaction on the seismic responses of a nuclear stack building

    International Nuclear Information System (INIS)

    Kabir, A.F.; Malik, L.E.; Maryak, M.E.

    1991-01-01

    A nuclear exhaust stack building was analyzed considering flexibility of the base mat and through-soil coupling with a nearby massive reactor building. The analysis indicated that the base mat flexibility and the proximity of the reactor building significantly affect the seismic responses of the stack building

  5. EPR by Areva. EPR the 1600+ MWe reactor

    International Nuclear Information System (INIS)

    2008-01-01

    This brochure presents the GEN III+ EPR reactor designed by the Areva and Siemens consortium. The EPR reactor is a direct descendent of the well-proven N4 and KONVOI reactors, the most modern reactors in France and Germany. The EPR was designed by teams from KWU/Siemens and Framatome, EDF in France and the major German utilities, working in collaboration with both French and German safety authorities. The EPR integrates the results of decades of R and D programs, in particular those performed by the CEA (French Atomic Energy Commission) and the Karlsruhe Research Center in Germany. The EPR benefits from the experience of several thousand reactor-years of operation of pressurized water reactor technology. This experience has put 87 AREVA PWRs online throughout the world. Innovative Features: - An outer shell covering the reactor building, the spent fuel building and two of the four safeguard buildings provides protection against large commercial or military aircraft crash. - A heavy neutron reflector that surrounds the reactor core lowers uranium consumption. - An axial economizer inside the steam generator allows a high level of steam pressure and therefore high plant efficiency. - A core catcher allows passive collection and retention of the molten core should the reactor vessel fail in the highly unlikely event of a core melt. - A digital technology and a fully computerized control room with an operator friendly man-machine interface improve the reactor protection system

  6. Nuclear research reactors

    International Nuclear Information System (INIS)

    1985-01-01

    It's presented data about nuclear research reactors in the world, retrieved from the Sien (Nuclear and Energetic Information System) data bank. The information are organized in table forms as follows: research reactors by countries; research reactors by type; research reactors by fuel and research reactors by purpose. (E.G.) [pt

  7. SOLASE: a conceptual laser fusion reactor design

    International Nuclear Information System (INIS)

    Conn, R.W.; Abdel-Khalik, S.I.; Moses, G.A.

    1977-12-01

    The SOLASE conceptual laser fusion reactor has been designed to elucidate the technological problems posed by inertial confinement fusion ractors. This report contains a detailed description of all aspects of the study including the physics of pellet implosion and burn, optics and target illumination, last mirror design, laser system analysis, cavity design, pellet fabrication and delivery, vacuum system requirements, blanket design, thermal hydraulics, tritium analysis, neutronics calculations, radiation effects, stress analysis, shield design, reactor and plant building layout, maintenance procedures, and power cycle design. The reactor is designed as a 1000 MW/sub e/ unit for central station electric power generation

  8. Safety design of prototype fast breeder reactor

    International Nuclear Information System (INIS)

    Bhoje, S.B.; Chetal, S.C.; Singh, Om Pal

    2004-01-01

    The basic design and safety design of Prototype Fast Breeder Reactor (PFBR) is presented. Design aspects covered include safety classification, seismic categorization, design basis conditions, design safety limits, core physics, core monitoring, shutdown system, decay heat removal system, protection against sodium leaks and tube leaks in steam generator, plant layout, radiation protection, event analysis, beyond design basis accidents, integrity of primary containment, reactor containment building and design pressure resulting from core disruptive accident. The measures provided in the design represent a robust case of the safety of the reactor. (authors)

  9. Nuclear reactor physics course for reactor operators

    International Nuclear Information System (INIS)

    Baeten, P.

    2006-01-01

    The education and training of nuclear reactor operators is important to guarantee the safe operation of present and future nuclear reactors. Therefore, a course on basic 'Nuclear reactor physics' in the initial and continuous training of reactor operators has proven to be indispensable. In most countries, such training also results from the direct request from the safety authorities to assure the high level of competence of the staff in nuclear reactors. The aim of the basic course on 'Nuclear Reactor Physics for reactor operators' is to provide the reactor operators with a basic understanding of the main concepts relevant to nuclear reactors. Seen the education level of the participants, mathematical derivations are simplified and reduced to a minimum, but not completely eliminated

  10. Building America

    Energy Technology Data Exchange (ETDEWEB)

    Brad Oberg

    2010-12-31

    IBACOS researched the constructability and viability issues of using high performance windows as one component of a larger approach to building houses that achieve the Building America 70% energy savings target.

  11. Building calculations

    DEFF Research Database (Denmark)

    Jensen, Bjarne Christian; Hansen, Svend Ole

    Textbook on design of large panel building including rules on robustness and a method for producing the Statical documentattion......Textbook on design of large panel building including rules on robustness and a method for producing the Statical documentattion...

  12. NASA Reactor Facility Hazards Summary. Volume 1

    Science.gov (United States)

    1959-01-01

    The Lewis Research Center of the National Aeronautics and Space Administration proposes to build a nuclear research reactor which will be located in the Plum Brook Ordnance Works near Sandusky, Ohio. The purpose of this report is to inform the Advisory Committee on Reactor Safeguards of the U. S. Atomic Energy Commission in regard to the design Lq of the reactor facility, the characteristics of the site, and the hazards of operation at this location. The purpose of this research reactor is to make pumped loop studies of aircraft reactor fuel elements and other reactor components, radiation effects studies on aircraft reactor materials and equipment, shielding studies, and nuclear and solid state physics experiments. The reactor is light water cooled and moderated of the MTR-type with a primary beryllium reflector and a secondary water reflector. The core initially will be a 3 by 9 array of MTR-type fuel elements and is designed for operation up to a power of 60 megawatts. The reactor facility is described in general terms. This is followed by a discussion of the nuclear characteristics and performance of the reactor. Then details of the reactor control system are discussed. A summary of the site characteristics is then presented followed by a discussion of the larger type of experiments which may eventually be operated in this facility. The considerations for normal operation are concluded with a proposed method of handling fuel elements and radioactive wastes. The potential hazards involved with failures or malfunctions of this facility are considered in some detail. These are examined first from the standpoint of preventing them or minimizing their effects and second from the standpoint of what effect they might have on the reactor facility staff and the surrounding population. The most essential feature of the design for location at the proposed site is containment of the maximum credible accident.

  13. Higher power density TRIGA research reactors

    International Nuclear Information System (INIS)

    Whittemore, W.L.

    2008-01-01

    The uranium zirconium hydride (U-ZrH) fuel is the fundamental feature of the TRIGA family of reactors that accounts for its widely recognized safety, good performance, economy of operation, and its acceptance worldwide. Of the 65 TRIGA reactors or TRIGA fueled reactors, several are located in hospitals or hospital complexes and in buildings that house university classrooms. These examples are a tribute to the high degree of safety of the operating TRIGA reactor. In the early days, the majority of the TRIGA reactors had power levels in the range from 10 to 250 kW, many with pulsing capability. An additional number had power levels up to 1 MW. By the late 1970's, seven TRIGA reactors with power levels up to 2 MW had been installed. A reduction in the rate of worldwide construction of new research reactors set in during the mid 1970's but construction of occasional research reactors has continued until the present. Performance of higher power TRIGA reactors are presented as well as the operation of higher power density reactor cores. The extremely safe TRIGA fuel, including the more recent TRIGA LEU fuel, offers a wide range of possible reactor configurations. A long core life is assured through the use of a burnable poison in the TRIGA LEU fuel. In those instances where large neutron fluxes are desired but relatively low power levels are also desired, the 19-rod hexagonal array of small diameter fuel rods offers exciting possibilities. The small diameter fuel rods have provided extremely long and trouble-free operation in the Romanian 14 MW TRIGA reactor

  14. Solar building

    OpenAIRE

    Zhang, Luxin

    2014-01-01

    In my thesis I describe the utilization of solar energy and solar energy with building integration. In introduction it is also mentioned how the solar building works, trying to make more people understand and accept the solar building. The thesis introduces different types of solar heat collectors. I compared the difference two operation modes of solar water heating system and created examples of solar water system selection. I also introduced other solar building applications. It is conv...

  15. Photocatalytic reactor

    Science.gov (United States)

    Bischoff, Brian L.; Fain, Douglas E.; Stockdale, John A. D.

    1999-01-01

    A photocatalytic reactor 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.

  16. Nuclear reactor

    International Nuclear Information System (INIS)

    Anthony, A.J.; Gruber, E.A.

    1979-01-01

    A nuclear reactor with control rods in channels between fuel assemblies wherein the fuel assemblies incorporate guide rods which protrude outwardly into the control rod channels to prevent the control rods from engaging the fuel elements. The guide rods also extend back into the fuel assembly such that they are relatively rigid members. The guide rods are tied to the fuel assembly end or support plates and serve as structural members which are supported independently of the fuel element. Fuel element spacing and support means may be attached to the guide rods. 9 claims

  17. Building 2000

    International Nuclear Information System (INIS)

    Den Ouden, C.; Steemers, T.C.

    1992-01-01

    This is the first volume of Building 2000, a pilot project of the Commission's R and D-programme 'Solar Energy Applications to Buildings' with the purpose of encouraging the adoption of solar architecture in large buildings. In this first rich illustrated volume the results of the design studies illustrating passive solar architecture in buildings in the European Community are presented in particular for the building categories as mentioned in the subtitle. In a second volume, a similar series of studies is presented for the building categories: office buildings, public buildings and hotels and holiday complexes. Several Design Support Workshops were organized during the Building 2000 programme during which Building 2000 design teams could directly exchange ideas with the various design advice experts represented at these workshops. In the second part of the Building 2000 final report a summary of a selection of many reports is presented (15 papers), as produced by Design Support experts. Most of the design support activities resulted in changes of the various designs, as have been reported by the design teams in the brochures presented in the first part of this book. It is to be expected that design aids and simulation tools for passive solar options, daylighting concepts, comfort criteria etc., will be utilized more frequently in the future. This will result in a better exchange of information between the actual design practitioners and the European R and D community. This technology transfer will result in buildings with a higher quality with respect to energy and environmental issues

  18. Auxiliary buildings

    International Nuclear Information System (INIS)

    Lakner, I.; Lestyan, E.

    1979-01-01

    The nuclear power station represents a complicated and a particular industrial project. Consequently, the design of the auxiliary buildings serving the power station (offices, kitchen, refreshment room, workshops, depots, water treatment plant building, boiler houses, etc.) requires more attention than usual. This chapter gives a short survey of the auxiliary buildings already completed and discusses the problems of their design, location and structure. (author)

  19. D and DR Reactors

    Data.gov (United States)

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

  20. Reactor core fuel management

    International Nuclear Information System (INIS)

    Silvennoinen, P.

    1976-01-01

    The subject is covered in chapters, entitled: concepts of reactor physics; neutron diffusion; core heat transfer; reactivity; reactor operation; variables of core management; computer code modules; alternative reactor concepts; methods of optimization; general system aspects. (U.K.)

  1. Hybrid adsorptive membrane reactor

    Science.gov (United States)

    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.

  2. Reactor container

    International Nuclear Information System (INIS)

    Oikawa, Hirohide; Otonari, Jun-ichiro; Tozaki, Yuka.

    1993-01-01

    Partition walls are disposed between a reactor pressure vessel and a suppression chamber to separate a dry well to an upper portion and a lower portion. A communication pipe is disposed to the partition walls. One end of the communication pipe is opened in an upper portion of the dry well at a position higher than a hole disposed to a bent tube of the suppression chamber. When coolants overflow from a depressurization valve by an erroneous operation of an emergency reactor core cooling device, the coolants accumulate in the upper portion of the dry well. When the pipeline is ruptured at the upper portion of the pressure vessel, only the inside of the pressure vessel and the upper portion of the dry well are submerged in water. In this case, the water level of the coolants does not elevate to the opening of the commuication pipe but they flow into the suppression chamber from the hole disposed to the bent tube. Since the coolants do not flow out to the lower portion of the dry well, important equipments such as control rod drives disposed at the lower portion of the dry wall can be prevented from submerging in water. (I.N.)

  3. Reactor container

    Energy Technology Data Exchange (ETDEWEB)

    Oikawa, Hirohide; Otonari, Jun-ichiro; Tozaki, Yuka.

    1993-09-07

    Partition walls are disposed between a reactor pressure vessel and a suppression chamber to separate a dry well to an upper portion and a lower portion. A communication pipe is disposed to the partition walls. One end of the communication pipe is opened in an upper portion of the dry well at a position higher than a hole disposed to a bent tube of the suppression chamber. When coolants overflow from a depressurization valve by an erroneous operation of an emergency reactor core cooling device, the coolants accumulate in the upper portion of the dry well. When the pipeline is ruptured at the upper portion of the pressure vessel, only the inside of the pressure vessel and the upper portion of the dry well are submerged in water. In this case, the water level of the coolants does not elevate to the opening of the commuication pipe but they flow into the suppression chamber from the hole disposed to the bent tube. Since the coolants do not flow out to the lower portion of the dry well, important equipments such as control rod drives disposed at the lower portion of the dry wall can be prevented from submerging in water. (I.N.).

  4. Reactor monitor

    International Nuclear Information System (INIS)

    Takada, Tamotsu.

    1992-01-01

    The device of the present invention monitors a reactor so that each of the operations for the relocation of fuel assemblies and the withdrawal and the insertion of control rods upon exchange of fuel assemblies and control rods in the reactor. That is, when an operator conducts relocating operation by way of a fuel assembly operation section, the device of the present invention judges whether the operation indication is adequate or not, based on the information of control rod arrangement in a control rod memory section. When the operation indication is wrong, a stop signal is sent to a fuel assembly relocating device. Further, when the operator conducts control rod operation by way of a control rod operation section, the device of the present invention judges in the control rod withdrawal judging section, as to whether the operation indication given by the operator is adequate or not by comparing it with fuel assembly arrangement information. When the operation indication is wrong, a stop signal is sent to control rod drives. With such procedures, increase of nuclear heating upon occurrence of erroneous operation can be prevented. (I.S.)

  5. Analysis of Coolant Options for Advanced Metal Cooled Nuclear Reactors

    National Research Council Canada - National Science Library

    Can, Levent

    2006-01-01

    .... The overall focus of this study is the build up of induced radioactivity in the coolant of metal cooled reactors as well as the evaluation of other physical and chemical properties of such coolants...

  6. Cable condition monitoring in a pressurized water reactor environment

    International Nuclear Information System (INIS)

    Al-Hussaini, T.J.

    1988-01-01

    Oconee Nuclear Station is the first nuclear plant designed, engineered and constructed by Duke Power Company. Even though the accelerated aging method was available to determine the life expectancy of the cable used in the reactor building, no natural aging data was available at that time. In order to be able to verify the condition of the reactor building cable over the life of the plant, an on-going cable monitoring plan was instituted. Various types of cable were selected to be monitored, and they were installed in cable life evaluation circuits in the reactor building. At five year intervals over the life of the plant, cable samples would be removed from these cable life evaluation circuits and tested to determine the effects of the reactor building environment on the integrity of the cable. A review of the cable life evaluation circuits and the results of the evaluation program to date is presented

  7. Nuclear Power: Outlook for New U.S. Reactors

    National Research Council Canada - National Science Library

    Parker, Larry; Holt, Mark

    2007-01-01

    Nearly three decades after the most recent order was placed for a new nuclear power plant in the United States, several utilities are now expressing interest in building a total of up to 30 new reactors...

  8. Flash Cracking Reactor for Waste Plastic Processing, Phase I

    Data.gov (United States)

    National Aeronautics and Space Administration — We propose to design, model, build, and test a novel flash cracking reactor to convert plastic waste, and potentially other unconventional hydrocarbon feedstocks,...

  9. Antineutrino monitoring for the Iranian heavy water reactor

    OpenAIRE

    Christensen, Eric; Huber, Patrick; Jaffke, Patrick; Shea, Thomas

    2014-01-01

    In this note we discuss the potential application of antineutrino monitoring to the Iranian heavy water reactor at Arak, the IR-40, as a non-proliferation measure. We demonstrate that an above ground detector positioned right outside the IR-40 reactor building could meet and in some cases significantly exceed the verification goals identified by IAEA for plutonium production or diversion from declared inventories. In addition to monitoring the reactor during operation, observing antineutrino ...

  10. Nuclear Reactor RA Safety Report, Vol. 2, Site characterization

    International Nuclear Information System (INIS)

    1986-11-01

    RA research reactor is located in the village Vinca, in the vicinity of the Danube right bank, downstream from Belgrade, about 12 km from the city center. The reactor building is in the shallow valley on a plateau about 100 m above the sea level and 20 m above the Danube level. Apart from geographic and demographic data of the reactor site this document covers geologic, seismology and hydrology data. Climatic, ecology and radioactivity data of the surroundings are included as well [sr

  11. Thermal-hydraulic analysis of the Three Mile Island Unit 2 reactor accident with THALES code

    International Nuclear Information System (INIS)

    Hashimoto, Kazuichiro; Soda, Kunihisa

    1991-10-01

    The OECD Nuclear Energy Agency (NEA) has established a Task Group in the Committee on the Safety of Nuclear Installations (CSNI) to perform an analysis of Three Mile Island Unit 2 (TMI-2) accident as a standard problem to benchmark severe accident computer codes and to assess the capability of the codes. The TMI-2 Analysis Exercise was performed at the Japan Atomic Energy Research Institute (JAERI) using the THALES (Thermal-Hydraulic Analysis of Loss-of-Coolant, Emergency Core Cooling and Severe Core Damage) - PM1/TMI code. The purpose of the analysis is to verify the capability of THALES-PM1/TMI code to describe accident progression in the actual plant. The present paper describes the final result of the TMI-2 Analysis Exercise performed at JAERI. (author)

  12. Nuclear reactor

    International Nuclear Information System (INIS)

    Schabert, H.P.; Weber, R.; Bauer, A.

    1975-01-01

    The refuelling of a PWR power reactor of about 1,200 MWe is performed by a transport pipe in the containment leading from an external to an internal fuel pit. A wagon to transport the fuel elements can go from a vertical loading position to an also vertical deloading position in the inner fuel pit via guide rollers. The necessary horizontal movement is effected by means of a cable line through the transport pipe which is inclined at least 10 0 . Gravity thus helps in the movement to the deloading position. The cable line with winch is fastened outside the containment. Swivelling devices tip the wagon from the horizontal to the vertical position or vice versa. Loading and deloading are done laterally. (TK/LH) [de

  13. NEUTRONIC REACTOR

    Science.gov (United States)

    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.

  14. Nuclear reactor

    International Nuclear Information System (INIS)

    Schweiger, F.; Glahe, E.

    1976-01-01

    In a nuclear reactor of the kind which is charged with spherical reaction elements and in which control rods are arranged to be thrust directly into the charge, each control rod has at least one screw thread on its external surface so that as the rod is thrust into the charge it is caused to rotate and thus make penetration easier. The length of each control rod may have two distinct portions, a latter portion which carries a screw thread and a lead-in portion which is shorter than the latter portion and which may carry a thread of greater pitch than that on the latter portion or may have a number of axially extending ribs instead of a thread

  15. Nuclear reactor neutron shielding

    Science.gov (United States)

    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.

  16. Field measurements and interpretation of TMI-2 instrumentation: IC-10-dPT

    International Nuclear Information System (INIS)

    Jones, J.E.; Smith, J.T.; Mathis, M.V.

    1982-01-01

    This report describes the measurements and results of the Control Rod Drive Bypass Flow IC-10-dPT. This instrument consists of a Bailey Type BY Process Computer Transmitter connected to a readout module by approximately 500 feet of cable through a penetration junction and an instrument mounting junction. The status of this instrument is uncertain, but it was producing a reasonable output reading of zero flow which could indicate it had not failed. As a result, measurements on this instrument were designed to determine if it were properly functioning

  17. Post TMI-2 view on the responsibilities of nuclear engineering educators

    International Nuclear Information System (INIS)

    Long, R.L.

    1980-01-01

    The Three Mile Island (TMI) accident of March 28, 1979 was the result of a complex set of interactions involving design deficiencies, equipment failure and human error. Nuclear engineering educators may need to accept responsibility for some of the underlying, industry-wide causes leading to the event. The many detailed investigations and recommendations following the accident are certain to have a significant impact on nuclear engineering education. Areas of impact include changes in curricula, increased demand for graduates, heavier involvement in utility staff training and education, and new approaches to university, industry, and societal interactions

  18. TMI-2: Lessons learned by the US Department of Energy: A programmatic perspective

    International Nuclear Information System (INIS)

    Schmitt, R.C.; Reno, H.W.; Bentley, K.J.; Owens, D.E.

    1990-03-01

    This report is a summary of the lessons learned by the US Department of Energy during its decade-long participation in the research and accident cleanup project at Three Mile Island Nuclear Power Station Unit 2 near Harrisburg, Pennsylvania. It is based on a review of a wide range of project documents and interviews with personnel from the many organizations involved. The lessons are organized into major subjects with a brief background section to orient the reader to that subject. The subjects are divided into sub-topics, each with a brief discussion and a series of lessons learned. The lessons are very brief and each is preceded with a keyword phrase to highlight its specific topic. References are given so that the details of the subject and the lesson can be further investigated. 99 refs., 24 figs

  19. Possible options for reducing occupational dose from the TMI-2 basement

    International Nuclear Information System (INIS)

    Munson, L.F.; Harty, R.

    1985-11-01

    The major sources of exposure in the basement include the enclosed stairwell/elevator shaft structure, water and sludge in the elevator shaft, cast concrete walls, concrete floor slab, water and sludge on the floor, and activity in the paint and loose surface contamination. The sources were identified using data obtained by the utility from water processing, water and solid samples, remote video inspections and radiation monitoring with a robot, and strings of thermoluminescent dosimeters lowered from upper elevations. The area dose rates in the basement range from approximately 4 R/hr (in the NE quadrant) to over 1100 R/hr (near the enclosed stairwell/elevator shaft structure). It is estimated that the basement contains between 11,000 and 21,000 curies of 137 Cs. Specific decontamination and cleanup techniques are discussed. These techniques include flushing with water, high-pressure water blasting, leaching, scabbling and chemical cleaning. The applicability of these techniques to the major sources of radiation are discussed, and possible approaches and work sequences for basement cleanup are given

  20. NRC action plan developed as a result of the TMI-2 accident. Volume 2

    International Nuclear Information System (INIS)

    1980-05-01

    The Action Plan provides a comprehensive and integrated plan for all actions judged necessary by the Nuclear Regulatory Commission to correct or improve the regulation and operation of nuclear facilities based on the experience from the accident at the Three Mile Island, Unit 2, nuclear facility and the official studies and investigations of the accident. The tables included in this volume list the recommendations from the various organizations and task forces investigating the accident at Three Mile Island. The tables are annotated to provide easy references to the associated parts of the Action Plan in Volume 1. The tables are also annotated to provide a shorthand indication of how the various recommendations are treated in the Action Plan

  1. Potential impact of licensee default on cleanup of TMI-2. Technical report

    International Nuclear Information System (INIS)

    Roberts, J.O.; Saltzman, J.

    1980-11-01

    Financial repercussions of the accident at Three Mile Island Unit 2 on the ability of the Licensee, Metropolitan Edison Co., to complete cleanup of the facility are examined. Potential impacts of licensee default on cleanup and alternatives to minimize the potential of bankruptcy are discussed. Specific recommendations are made regarding steps the Nuclear Regulatory Commission might take in keeping with its regulatory functions and its mission to protect the public health and safety

  2. TMI-2 [Three-Mile Island-Unit 2] rail cask and railcar maintenance

    International Nuclear Information System (INIS)

    Tyacke, M.J.; Ayers, A.L. Jr.; Ball, L.J.; Anselmo, A.A.

    1988-02-01

    This paper describes the NuPac 125-B cask system (i.e., cask and railcar), and the maintenance and inspection requirements for that system. The paper discusses the operations being done to satisfy those requirements and how, in some cases, it has been efficient for the operations to be more rigorous than the requirements. Finally, this paper discusses the experiences gained from those operations and how specific hardware and procedural enhancements have resulted in a reliable and continuous shipping campaign. 2 refs., 2 figs

  3. Nondestructive techniques for assaying fuel debris in piping at Three Mile Island Unit 2

    International Nuclear Information System (INIS)

    Vinjamuri, K.; McIsaac, C.V.; Beller, L.S.; Isaacson, L.; Mandler, J.W.; Hobbins, R.R. Jr.

    1981-11-01

    Four major categories of nondestructive techniques - ultrasonic, passive gamma ray, infrared detection, and remote video examination - have been determined to be feasible for assaying fuel debris in the primary coolant system of the Three Mile Island Unit 2 (TMI-2) Reactor. Passive gamma ray detection is the most suitable technique for the TMI-2 piping; however, further development of this technique is needed for specific application to TMI-2

  4. Building envelope

    CSIR Research Space (South Africa)

    Gibberd, Jeremy T

    2009-01-01

    Full Text Available have controllable internal blinds and external solar shading which can be used to maximise internal daylight quality and avoid glare and solar gain. • Ecological: Green building envelopes aim to support the development of ecosystems and plant... problems such as large solar heat gains which can only be solved through mechanical plant. • Large sections: 1:10 sections of building envelopes are very useful in understanding the design of a building envelope and should be developed early...

  5. Laboratory Building

    Energy Technology Data Exchange (ETDEWEB)

    Herrera, Joshua M. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States)

    2015-03-01

    This report is an analysis of the means of egress and life safety requirements for the laboratory building. The building is located at Sandia National Laboratories (SNL) in Albuquerque, NM. The report includes a prescriptive-based analysis as well as a performance-based analysis. Following the analysis are appendices which contain maps of the laboratory building used throughout the analysis. The top of all the maps is assumed to be north.

  6. Reactor core and passive safety systems descriptions of a next generation pressure tube reactor - mechanical aspects

    Energy Technology Data Exchange (ETDEWEB)

    Yetisir, M.; Gaudet, M.; Rhodes, D.; Hamilton, H.; Pencer, J. [Atomic Energy of Canada Limited, Chalk River, ON (Canada)

    2014-07-01

    Canada has been developing a channel-type supercritical water-cooled nuclear reactor concept, often called the Canadian SCWR. The objective of this reactor concept is to meet the technology goals of the Generation IV International Forum (GIF) for the next generation nuclear reactor development, which include enhanced safety features (inherent safe operation and deploying passive safety features), improved resource utilization, sustainable fuel cycle, and greater proliferation resistance than Generation III nuclear reactors. The Canadian SCWR core concept consists of a high-pressure inlet plenum, a separate low-pressure heavy water moderator contained in a calandria vessel, and 336 pressure tubes surrounded by the moderator. The reactor uses supercritical water as a coolant, and a direct steam power cycle to generate electricity. The reactor concept incorporates advanced safety features such as passive core cooling, long-term decay heat rejection to the environment and fuel melt prevention via passive moderator cooling. These features significantly reduce core damage frequency relative to existing nuclear reactors. This paper presents a description of the design concepts for the Canadian SCWR core, reactor building layout and the plant layout. Passive safety concepts are also described that address containment and core cooling following a loss-of coolant accident, as well as long term reactor heat removal at station blackout conditions. (author)

  7. The Effect Of Beryllium Interaction With Fast Neutrons On the Reactivity Of ETRR-2 Research Reactor

    International Nuclear Information System (INIS)

    Aziz, M.; El Messiry, A.M.

    2000-01-01

    The effect of beryllium interactions with fast neutrons is studied for Etrr 2 research reactors. Isotope build up inside beryllium blocks is calculated under different irradiation times. a new model for the Etrr 2 research reactor is designed using MCNP code to calculate the reactivity and flux change of the reactor due to beryllium poison

  8. Department of reactor technology

    International Nuclear Information System (INIS)

    1980-01-01

    The activities of the Department of Reactor Technology at Risoe during 1979 are described. The work is presented in five chapters: Reactor Engineering, Reactor Physics and Dynamics, Heat Transfer and Hydraulics, The DR 1 Reactor, and Non-Nuclear Activities. A list of the staff and of publications is included. (author)

  9. Survey of research reactors

    International Nuclear Information System (INIS)

    Boek, H.; Villa, M.

    2004-06-01

    A survey of reasearch reactors based on the IAEA Nuclear Research Reactor Data Base (RRDB) was done. This database includes information on 273 operating research reactors ranging in power from zero to several hundred MW. From these 273 operating research reactors 205 reactors have a power level below 5 MW, the remaining 68 reactors range from 5 MW up to several 100 MW thermal power. The major reactor types with common design are: Siemens Unterrichtsreaktors, 1.2 Argonaut reactors, Slowpoke reactors, the miniature neutron source reactors, TRIGA reactors, material testing reactors and high flux reactors. Technical data such as: power, fuel material, fuel type, enrichment, maximum neutron flux density and experimental facilities for each reactor type as well as a description of their utilization in physics and chemistry, medicine and biology, academic research and teaching, training purposes (students and physicists, operating personnel), industrial application (neutron radiography, silicon neutron transmutation doping facilities) are provided. The geographically distribution of these reactors is also shown. As conclusions the author discussed the advantages (low capital cost, low operating cost, low burn up, simple to operate, safe, less restrictive containment and sitting requirements, versatility) and disadvantages (lower sensitivity for NAA, limited radioisotope production, limited use of neutron beams, limited access to the core, licensing) of low power research reactors. 24 figs., refs. 15, Tab. 1 (nevyjel)

  10. FBR type reactor

    International Nuclear Information System (INIS)

    Kimura, Kimitaka; Fukuie, Ken; Iijima, Tooru; Shimpo, Masakazu.

    1994-01-01

    In an FBR type reactor for exchanging fuels by pulling up reactor core upper mechanisms, a connection mechanism is disposed for connecting the top of the reactor core and the lower end of the reactor core upper mechanisms. In addition, a cylindrical body is disposed surrounding the reactor core upper mechanisms, and a support member is disposed to the cylindrical body for supporting an intermediate portion of the reactor core upper mechanisms. Then, the lower end of the reactor core upper mechanisms is connected to the top of the reactor core. Same displacements are caused to both of them upon occurrence of earthquakes and, as a result, it is possible to eliminate mutual horizontal displacement between a control rod guide hole of the reactor core upper mechanisms and a control rod insertion hole of the reactor core. In addition, since the intermediate portion of the reactor core upper mechanisms is supported by the support member disposed to the cylindrical body surrounding the reactor core upper mechanisms, deformation caused to the lower end of the reactor core upper mechanisms is reduced, so that the mutual horizontal displacement with respect to the control rod insertion hole of the reactor core can be reduced. As a result, performance of control rod insertion upon occurrence of the earthquakes is improved, so that reactor shutdown is conducted more reliably to improve reactor safety. (N.H.)

  11. Fast breeder reactors

    International Nuclear Information System (INIS)

    Heinzel, V.

    1975-01-01

    The author gives a survey of 'fast breeder reactors'. In detail the process of breeding, the reasons for the development of fast breeders, the possible breeder reactors, the design criteria, fuels, cladding, coolant, and safety aspects are reported on. Design data of some experimental reactors already in operation are summarized in stabular form. 300 MWe Prototype-Reactors SNR-300 and PFR are explained in detail and data of KWU helium-cooled fast breeder reactors are given. (HR) [de

  12. Determining Reactor Neutrino Flux

    OpenAIRE

    Cao, Jun

    2011-01-01

    Flux is an important source of uncertainties for a reactor neutrino experiment. It is determined from thermal power measurements, reactor core simulation, and knowledge of neutrino spectra of fuel isotopes. Past reactor neutrino experiments have determined the flux to (2-3)% precision. Precision measurements of mixing angle $\\theta_{13}$ by reactor neutrino experiments in the coming years will use near-far detector configurations. Most uncertainties from reactor will be canceled out. Understa...

  13. Nuclear reactor coolant channels

    International Nuclear Information System (INIS)

    Macbeth, R.V.

    1978-01-01

    A nuclear reactor coolant channel is described that is suitable for sub-cooled reactors as in pressurised water reactors as well as for bulk boiling, as in boiling water reactors and steam generating nuclear reactors. The arrangement aims to improve heat transfer between the fuel elements and the coolant. Full constructional details are given. See also other similar patents by the author. (U.K.)

  14. Scenario for commercialization of fast breeder reactors

    International Nuclear Information System (INIS)

    Kumaoka, Yoshio; Sato, Morihiko

    1989-01-01

    To realize the commercialization of fast breeder reactors (FBRs), it is essential to reduce construction costs to the same level as those for the current light water reactors. For this target to be attained, a highly important factor is to reduce to the lowest-levels possible the quantities of materials and volume of the buildings required for the primary and secondary sodium loops of the FBR. In this direction, an innovative compact FBR plant concept which holds promise for commercialization has been developed by introducing the pooltype reactor concept with the shortest possible secondary sodium loops, realized by coupling electromagnetic pumps with the steam generators. In comparison with the French Super Phenix reactor, for example, the construction of this 1,300-MWe FBR plant could be achieved with half the material quantities and plant volume required by the former type. (author)

  15. Unconventional liquid metal cooled fast reactors

    International Nuclear Information System (INIS)

    Spinrad, B.I.; Rohach, A.F.; Razzaque, M.M.

    1989-06-01

    This report describes the rationale for, design of and analytical studies on an unconventional sodium-cooled power reactor, called the Trench Reactor. It derives its name from the long, narrow sodium pool in which the reactor is placed. Unconventional features include: pool shape; reactor shape (also long and narrow); reflector control; low power density; hot-leg primary pumping; absence of a cold sodium pool; large core boxes rather than a large number of subassemblies; large diameter metal fuel; vessel suspension from cables; and vessel cooling by natural circulation of building atmosphere (nitrogen) at all times. These features all seem feasible. They result in a system that is capable of at least a ten year reload interval and shows good safety through direct physical response to loss-of-heat-sink, loss-of-flow and limited-reactivity nuclear transients. 43 figs., 43 tabs

  16. Space reactors, a prospective for the future

    International Nuclear Information System (INIS)

    Wahlquist, E.; Voss, S.S.

    1989-01-01

    The power requirements for future space missions are increasing and alternate power systems will be required to meet these needs. Therefore, in the early 1980's a tri-agency space reactor program, the SP-100, was initiated that is capable of meeting the higher power requirements. To understand the current space reactor program, it is important to review it in the context of past space nuclear programs - including radioisotopes, nuclear rockets and reactors. Initial effort on these programs began in the mid-1950's. Radioisotope generators have been flown on a variety of missions and are continuing to be used. The space reactor and nuclear rocket programs were technically successful but were both terminated in 1973. The current SP-100 program builds on those earlier programs

  17. Nuclear reactor

    International Nuclear Information System (INIS)

    Irion, L.; Tautz, J.; Ulrych, G.

    1976-01-01

    This additional patent complements the arrangement of non-return valves to prevent loss of cooling water on fracture of external tubes in the main coolant circuit (according to PS 24 24 427.7) by ensuring that the easily movable valves only operate in case of a fault, but do not flutter in operation, because the direction of flow is not the same at each location where they are installed. The remedy for this undesirable effect consists of allocating 1 non-return valve unit with 5 to 10 valves to each (of several) ducts for the cooling water intake. These units are installed in the annular space between the reactor vessel and the pressure vessel below the inlet of the ducts. Due to flow guidance surfaces in the same space, the incoming cooling water is deflected downwards and as the guiding surfaces are closed at the sides, must pass parallel to the valves of the non-return valve unit. On fracture of the external cooling water inlet pipe concerned, all valves of this unit close due to reversal of flow on the outlet side. (TK) [de

  18. Reactor container

    International Nuclear Information System (INIS)

    Kagawa, Tatsuo; Yanai, Ryoichi.

    1976-01-01

    Object: To provide a reactor container which is free from water shock action or condensing vibrations and cannot be readily broken by a missile from a pump impeller, pipe whipping, steam jet reaction, etc., and which also quickly condenses issuing steam and possesses a large vibration-proof strength. Structure: A high pressure containment vessel accommodating a pressure container includes a plurality of pressurized water tanks arranged along its inner periphery, and a pneumatic valve is provided in a lower portion of each of these pressurized water tanks. If an accident occurs, vapor is caused to issue from the pressure container into the vessel. When a certain value is reached, the pneumatic valves are opened, whereby the gas within the pressurized water tanks causes pressurized water to flow through the pipe and be ejected from spray nozzles to cause condensation of water within the vessel. Further, water of a pool within the container is circulated to allow heat release to the outside. (Horiuchi, T.)

  19. The energy gap and the fast reactor

    International Nuclear Information System (INIS)

    Hill, J.

    1977-01-01

    The background to the development of fast reactors is summarized. In Britain, the results of the many experiments performed, the operation of the Dounreay Fast Reactor for the past 18 years and the first year's operation of the larger Prototype Fast Reactor have all been very encouraging, in that they demonstrated that the performance corresponded well with predictions, breeding is possible, and the system is exceptionally stable in operation. The next step in fast reactor engineering is to build a full-scale fast reactor power station. There would seem to be little reason to expect more trouble than could reasonably be expected in constructing any large project of this general nature. However, from an engineering point of view continuity of experience is required. If a decision to build a commercial fast reactor were taken today there would be a 14-year gap between strating this and the start of the Prototype Fast Reactor. This is already much too long. From an environmental standpoint we have to demonstrate that we can manufacture and reprocess fast reacctor fuel for a substantial programme in a way that does not lead to pollution of the environment, and that plutonium-containing fuel can be transported in the quantities required in safety and in a way that does not attract terrorists or require a private army to ensure its security. Finally, we have to find a way to allow many countries to obtain the energy they need from fast reactors, without leading to the proliferation of nuclear weapons or weapons capability. (author)

  20. Design of a multipurpose research reactor

    International Nuclear Information System (INIS)

    Sanchez Rios, A.A.

    1990-01-01

    The availability of a research reactor is essential in any endeavor to improve the execution of a nuclear programme, since it is a very versatile tool which can make a decisive contribution to a country's scientific and technological development. Because of their design, however, many existing research reactors are poorly adapted to certain uses. In some nuclear research centres, especially in the advanced countries, changes have been made in the original designs or new research prototypes have been designed for specific purposes. These modifications have proven very costly and therefore beyond the reach of developing countries. For this reason, what the research institutes in such countries need is a single sufficiently versatile nuclear plant capable of meeting the requirements of a nuclear research programme at a reasonable cost. This is precisely what a multipurpose reactor does. The Mexican National Nuclear Research Institute (ININ) plans to design and build a multipurpose research reactor capable at the same time of being used for the development of reactor design skills and for testing nuclear materials and fuels, for radioisotopes production, for nuclear power studies and basic scientific research, for specialized training, and so on. For this design work on the ININ Multipurpose Research Reactor, collaborative relations have been established with various international organizations possessing experience in nuclear reactor design: Atomehnergoeksport of the USSR: Atomic Energy of Canada Limited (AECL); General Atomics (GA) of the USA; and Japan Atomic Energy Research Institute

  1. Conceptual study of fusion-driven transmutation reactor with ITER physics and engineering constraints

    Science.gov (United States)

    Hong, Bong

    2011-10-01

    A conceptual study of fusion-driven transmutation reactor was performed based on ITER physics and engineering constraints. A compact reactor concept is desirable from an economic viewpoint. For the optimal design of a reactor, a radial build of reactor components has to be determined by considering the plasma physics and engineering constraints which inter-relate various reactor components. In a transmutation reactor, design of blanket and shield play a key role in determining the size of a reactor; the blanket should produce enough tritium for tritium self-sufficiency, the transmutation rate of waste has to be maximized, and the shield should provide sufficient protection for the superconducting toroidal field (TF) coil. To determine the radial build of the blanket and the shield, not only a radiation transport analysis but also a burnup calculation were coupled with the system analysis and it allowed the self-consistent determination of the design parameters of a transmutation reactor.

  2. Reactor Physics Training

    International Nuclear Information System (INIS)

    Baeten, P.

    2007-01-01

    University courses in nuclear reactor physics at the universities consist of a theoretical description of the physics and technology of nuclear reactors. In order to demonstrate the basic concepts in reactor physics, training exercises in nuclear reactor installations are also desirable. Since the number of reactor facilities is however strongly decreasing in Europe, it becomes difficult to offer to students a means for demonstrating the basic concepts in reactor physics by performing training exercises in nuclear installations. Universities do not generally possess the capabilities for performing training exercises. Therefore, SCK-CEN offers universities the possibility to perform (on a commercial basis) training exercises at its infrastructure consisting of two research reactors (BR1 and VENUS). Besides the organisation of training exercises in the framework of university courses, SCK-CEN also organizes theoretical courses in reactor physics for the education and training of nuclear reactor operators. It is indeed a very important subject to guarantee the safe operation of present and future nuclear reactors. In this framework, an understanding of the fundamental principles of nuclear reactor physics is also necessary for reactor operators. Therefore, the organisation of a basic Nuclear reactor physics course at the level of reactor operators in the initial and continuous training of reactor operators has proven to be indispensable. In most countries, such training also results from the direct request from the safety authorities to assure the high level of competence of the staff in nuclear reactors. The objectives this activity are: (1) to provide training and education activities in reactor physics for university students and (2) to organise courses in nuclear reactor physics for reactor operators

  3. Dynamic response characteristics in the vertical direction of building structural systems

    International Nuclear Information System (INIS)

    Watabe, Makoto; Kubo, Tetsuo; Yamanouchi, Hiroyuki; Ohkawa, Izuru

    1985-01-01

    In nuclear reactor buildings, important systems such as generators or steam pipes are placed or fixed on the floor slabs of the building. To evaluate their dynamic responses during earthquakes, it is essential to obtain dynamic characteristics in the vertical direction of the reactor building. This study is to reveal the vertical responses of a building structural system by vibration tests with a shaker, and to verify the analytical computer program code for vertical responses. (author)

  4. Building Procurement

    DEFF Research Database (Denmark)

    Andersson, Niclas

    2007-01-01

    ‘The procurement of construction work is complex, and a successful outcome frequently elusive’. With this opening phrase of the book, the authors take on the challenging job of explaining the complexity of building procurement. Even though building procurement systems are, and will remain, complex...... despite this excellent book, the knowledge, expertise, well-articulated argument and collection of recent research efforts that are provided by the three authors will help to make project success less elusive. The book constitutes a thorough and comprehensive investigation of building procurement, which......, which gives the book a challenging contribution to the existing body of knowledge....

  5. The fast breeder reactor

    International Nuclear Information System (INIS)

    Collier, J.

    1990-01-01

    The arguments for and against the fast breeder reactor are debated. The case for the fast reactor is that the world energy demand will increase due to increasing population over the next forty years and that the damage to the global environment from burning fossil fuels which contribute to the greenhouse effect. Nuclear fission is the only large scale energy source which can achieve a cut in the use of carbon based fuels although energy conservation and renewable sources will also be important. Fast reactors produce more energy from uranium than other types of (thermal) reactors such as AGRs and PWRs. Fast reactors would be important from about 2020 onwards especially as by then many thermal reactors will need to be replaced. Fast reactors are also safer than normal reactors. The arguments against fast reactors are largely economic. The cost, especially the capital cost is very high. The viability of the technology is also questioned. (UK)

  6. Building of nuclear power plant

    International Nuclear Information System (INIS)

    Saito, Takashi.

    1997-01-01

    A first nuclear plant and a second nuclear power plant are disposed in adjacent with each other in a building for a nuclear reactor. A reactor container is disposed in each of the plants, and each reactor container is surrounded by a second containing facility. A repairing chamber capable of communicating with the secondary containing facilities for both of the secondary containing facilities is disposed being in contact with the second containing facility of each plant for repairing control rod driving mechanisms or reactor incorporated-type recycling pumps. Namely, the repairing chamber is in adjacent with the reactor containers of both plants, and situated between both of the plants as a repairing chamber to be used in common for both plants. Air tight inlet/exit doors are formed to the inlets/exits of both plants of the repairing chamber. Space for the repairing chamber can be reduced to about one half compared with a case where the repairing chamber is formed independently on each plant. (I.N.)

  7. Concept and designs of new-generation fast reactors

    International Nuclear Information System (INIS)

    Mitenkov, F.M.

    1993-01-01

    This article discusses the general safety requirements and characteristics for future nuclear power plants. It examines various designs - loop, block, and integrated layouts for reactors. Specifically, the article focuses an integrated design for sodium-cooled fast reactors noting that the BN-600 reactor has operated accident-free over the past 12 years. An obvious advantage of this scheme is that the coolant of the primary loop is localized in one volume (in a vessel), there are no short connections and large-diameter pipes, which of course sharply reduces the probability in coolant leaks. With an integrated scheme the problem of embrittlement of the reactor vessel by neutron irradiation is obviated. The neutron fluence for the vessels of the AST-500 and VPBER-600 reactors, built with an integrated scheme, is less than 10 17 cm -2 . Such a fluence does not cause any appreciable change in the mechanical properties of the vessel steel. The integrated layout of the reactor makes it possible to build a containment vessel. In this case it is possible to eliminate the danger of the reactor core drying out and thus cooling of the reactor in emergency situations can be simplified substantially. In an integrated layout, however, access is more difficult to the equipment inside the reactor, thus limiting or complicating maintenance work. The integrated layout, therefore, requires the use of highly reliable equipment built according to designs that have been proven in operation and have been passed representative service-life tests under laboratory conditions. The integrated layout considerably increases the mass and size characteristics of the reactor. New solutions thus are needed for the organization of work on reactor fabrication and assembly. In the case of the BN-600 and Superphenix reactors the welding of the reactor vessels and the assembly work were done on the building site

  8. Safeguarding research reactors

    International Nuclear Information System (INIS)

    Powers, J.A.

    1983-03-01

    The report is organized in four sections, including the introduction. The second section contains a discussion of the characteristics and attributes of research reactors important to safeguards. In this section, research reactors are described according to their power level, if greater than 25 thermal megawatts, or according to each fuel type. This descriptive discussion includes both reactor and reactor fuel information of a generic nature, according to the following categories. 1. Research reactors with more than 25 megawatts thermal power, 2. Plate fuelled reactors, 3. Assembly fuelled reactors. 4. Research reactors fuelled with individual rods. 5. Disk fuelled reactors, and 6. Research reactors fuelled with aqueous homogeneous fuel. The third section consists of a brief discussion of general IAEA safeguards as they apply to research reactors. This section is based on IAEA safeguards implementation documents and technical reports that are used to establish Agency-State agreements and facility attachments. The fourth and last section describes inspection activities at research reactors necessary to meet Agency objectives. The scope of the activities extends to both pre and post inspection as well as the on-site inspection and includes the examination of records and reports relative to reactor operation and to receipts, shipments and certain internal transfers, periodic verification of fresh fuel, spent fuel and core fuel, activities related to containment and surveillance, and other selected activities, depending on the reactor

  9. Nuclear reactor instrumentation at research reactor renewal

    International Nuclear Information System (INIS)

    Baers, B.; Pellionisz, P.

    1981-10-01

    The paper overviews the state-of-the-art of research reactor renewals. As a case study the instrumentation reconstruction of the Finnish 250 kW TRIGA reactor is described, with particular emphasis on the nuclear control instrumentation and equipment which has been developed and manufactured by the Central Research Institute for Physics, Budapest. Beside the presentation of the nuclear instrument family developed primarily for research reactor reconstructions, the quality assurance policy conducted during the manufacturing process is also discussed. (author)

  10. Physical protection of power reactors

    International Nuclear Information System (INIS)

    Darby, J.L.

    1979-01-01

    Sandia Laboratories has applied a systematic approach to designing physical protection systems for nuclear facilities to commercial light-water reactor power plants. A number of candidate physical protection systems were developed and evaluated. Focus is placed on the design of access control subsystems at each of three plant layers: the protected area perimeter, building surfaces, and vital areas. Access control refers to barriers, detectors, and entry control devices and procedures used to keep unauthorized personnel and contraband out of the plant, and to control authorized entry into vital areas within the plant

  11. Deep underground reactor (passive heat removal of LWR with hard neutron energy spectrum)

    International Nuclear Information System (INIS)

    Hiroshi, Takahashi

    2001-01-01

    To run a high conversion reactor with Pu-Th fueled tight fueled assembly which has a long burn-up of a fuel, the reactor should be sited deep underground. By putting the reactor deep underground heat can be removed passively not only during a steady-state run and also in an emergency case of loss of coolant and loss of on-site power; hence the safety of the reactor can be much improved. Also, the evacuation area around the reactor can be minimized, and the reactor placed near the consumer area. This approach reduces the cost of generating electricity by eliminating the container building and shortening transmission lines. (author)

  12. Reactor costs and maintenance, with reference to the Culham Mark II conceptual tokamak reactor design

    International Nuclear Information System (INIS)

    Hancox, R.; Mitchell, J.T.D.

    1977-01-01

    Published designs of tokamak reactors have proposed conceptual solutions for most of the technological problems encountered. Two areas which remain uncertain, however, are the capital cost of the reactor and the practicability of reactor maintenance. A cost estimate for the Culham Conceptual Tokamak Reactor (Mk I) is presented. The capital cost of a power station incorporating this reactor would be significantly higher than that of an equivalent fast breeder fission power station, mainly because of the low power density of the fusion reactor which affects both the reactor and building costs. To reduce the fusion station capital costs a new conceptual design is proposed (Mk II) which incorporates a shaped plasma cross-section to give a higher plasma pressure ratio, βsub(t) approximately 0.1. Since the higher power density implies more severe radiation damage of the blanket structure, the question of reactor maintenance assumes greater importance. With the proposed scheme for regular replacement of the blanket, a fusion power station availability around 0.9 should be achievable. (author)

  13. Building Data

    Data.gov (United States)

    Town of Cary, North Carolina — Explore real estate information about buildings in the Town of Cary.This file is created by the Town of Cary GIS Group. It contains data from both the Wake, Chatham...

  14. Tritium resources available for fusion reactors

    Science.gov (United States)

    Kovari, M.; Coleman, M.; Cristescu, I.; Smith, R.

    2018-02-01

    The tritium required for ITER will be supplied from the CANDU production in Ontario, but while Ontario may be able to supply 8 kg for a DEMO fusion reactor in the mid-2050s, it will not be able to provide 10 kg at any realistic starting time. The tritium required to start DEMO will depend on advances in plasma fuelling efficiency, burnup fraction, and tritium processing technology. It is in theory possible to start up a fusion reactor with little or no tritium, but at an estimated cost of 2 billion per kilogram of tritium saved, it is not economically sensible. Some heavy water reactor tritium production scenarios with varying degrees of optimism are presented, with the assumption that only Canada, the Republic of Korea, and Romania make tritium available to the fusion community. Results for the tritium available for DEMO in 2055 range from zero to 30 kg. CANDU and similar heavy water reactors could in theory generate additional tritium in a number of ways: (a) adjuster rods containing lithium could be used, giving 0.13 kg per year per reactor; (b) a fuel bundle with a burnable absorber has been designed for CANDU reactors, which might be adapted for tritium production; (c) tritium production could be increased by 0.05 kg per year per reactor by doping the moderator with lithium-6. If a fusion reactor is started up around 2055, governments in Canada, Argentina, China, India, South Korea and Romania will have the opportunity in the years leading up to that to take appropriate steps: (a) build, refurbish or upgrade tritium extraction facilities; (b) extend the lives of heavy water reactors, or build new ones; (c) reduce tritium sales; (d) boost tritium production in the remaining heavy water reactors. All of the alternative production methods considered have serious economic and regulatory drawbacks, and the risk of diversion of tritium or lithium-6 would also be a major concern. There are likely to be serious problems with supplying tritium for future

  15. Reactor Physics Programme

    International Nuclear Information System (INIS)

    De Raedt, C.

    2000-01-01

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

  16. Reactor Physics Programme

    Energy Technology Data Exchange (ETDEWEB)

    De Raedt, C

    2000-07-01

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

  17. Earthquake-proof supporting structure in reactor vessel

    International Nuclear Information System (INIS)

    Sakurai, Akio; Sekine, Katsuhisa; Madokoro, Manabu; Katoono, Shin-ichi; Konno, Mutsuo; Suzuki, Takuro.

    1990-01-01

    Conventional earthquake-proof structure comprises a vessel vibration stopper integrated to a reactor vessel, powder for restricting the horizontal displacements, a safety vessel surrounds the outer periphery of the reactor vessel and a safety vessel vibration stopper integrated therewith, which are fixed to buildings. However, there was a problem that a great amount of stresses are generated in the base of the reactor vessel vibration stopper due to reaction of the powders which restrict thermal expansion. In order to remarkably reduce the reaction of the powers, powders are charged into a spaces formed between each of the reactor vessel vibration stopper, the safety vessel vibration stopper and the flexible member disposed between them. According to this constitution, the reactor vessel vibration stopper does not undergo a great reaction of the powers upon thermal expansion of the reactor vessel to moderate the generated stresses, maintain the strength and provide earthquake-proof supporting function. (N.H.)

  18. Experimental reactors in the European Community and their utilization

    International Nuclear Information System (INIS)

    Ehringer, H.; Lecoq, J.P.

    1976-01-01

    Research and test reactors which in the first years of the peaceful use of nuclear energy had to found the basis for building and operation of commercial nuclear power plants, having achieved their aim, have faded into the background of the report. They still play an important role, however, for the further development of today's power reactor generation and for the development of progressing reactor lines as well as for fuel and material irradiation, for isotope production and, last but not least, for research and training. At the moment, over 100 test reactors are being operated in the widest sense in the European Community. In the present survey, their purpose and charge are dealt with particular consideration to the more important materials test reactors and to the programme reactors. (orig./LH) [de

  19. Reactor container and method of constructing the same

    International Nuclear Information System (INIS)

    Ichiki, Tadaharu; Muramatsu, Yutaka.

    1976-01-01

    Purpose: To reduce the construction term when constructing a reactor container and also alleviate the temperature requirements for the shielding wall by concurrently forwarding the work of constructing the reactor container and the shielding wall. Constitution: When constructing a reactor container, a shielding wall is constructed from lower layers on the outer side of the reactor container via a gap therefrom concurrently with the work of building the steel container. After the construction of the reactor container is completed, the reactor container is subjected to a pressure resistance test, leakage test, welding inspection and so forth. After the various tests and inspections are completed, concrete blocks as a protective material are stacked on the inner side of a steel frame mounted on the inner wall of the shielding wall, and they are secured to each other and also to the steel frame by concrete block holding members. (Moriyama, K.)

  20. Design project SP-RA, Construction specifications of the sanitary corridor - SP building

    International Nuclear Information System (INIS)

    Bon, B.

    1956-01-01

    According to the design task the sanitary corridor satisfies the needs of both RA reactor and HL (Hot laboratory) buildings. This building is located between RA and HL buildings and is connected to both by corridors 3,3 m wide. The access to the building id from the south, i.e. from the local road in the Vinca Institute [sr