WorldWideScience

Sample records for severe containment requirements

  1. Containment severe accident thermohydraulic phenomena

    International Nuclear Information System (INIS)

    Frid, W.

    1991-08-01

    This report describes and discusses the containment accident progression and the important severe accident containment thermohydraulic phenomena. The overall objective of the report is to provide a rather detailed presentation of the present status of phenomenological knowledge, including an account of relevant experimental investigations and to discuss, to some extent, the modelling approach used in the MAAP 3.0 computer code. The MAAP code has been used in Sweden as the main tool in the analysis of severe accidents. The dependence of the containment accident progression and containment phenomena on the initial conditions, which in turn are heavily dependent on the in-vessel accident progression and phenomena as well as associated uncertainties, is emphasized. The report is in three parts dealing with: * Swedish reactor containments, the severe accident mitigation programme in Sweden and containment accident progression in Swedish PWRs and BWRs as predicted by the MAAP 3.0 code. * Key non-energetic ex-vessel phenomena (melt fragmentation in water, melt quenching and coolability, core-concrete interaction and high temperature in containment). * Early containment threats due to energetic events (hydrogen combustion, high pressure melt ejection and direct containment heating, and ex-vessel steam explosions). The report concludes that our understanding of the containment severe accident progression and phenomena has improved very significantly over the parts ten years and, thereby, our ability to assess containment threats, to quantify uncertainties, and to interpret the results of experiments and computer code calculations have also increased. (au)

  2. Containment leakage rate testing requirements

    International Nuclear Information System (INIS)

    Arndt, E.G.

    1992-01-01

    This report presents the status of several documents under revision or development that provide requirements and guidance for testing nuclear power plant containment systems for leakage rates. These documents include the general revision to 10 CFR Part 50, Appendix J; the regulatory guide affiliated with the revision to Appendix J; the national standard that the regulatory guide endorses, ANSI/ANS-56.8, 'Containment System Leakage Rate Testing Requirements'; and the draft industry Licensing Topical Report, 'Standardized Program for Primary Containment Integrity Testing'. The actual or potential relationships between these documents are also explored

  3. Integrated severe accident containment analysis with the CONTAIN computer code

    International Nuclear Information System (INIS)

    Bergeron, K.D.; Williams, D.C.; Rexroth, P.E.; Tills, J.L.

    1985-12-01

    Analysis of physical and radiological conditions iunside the containment building during a severe (core-melt) nuclear reactor accident requires quantitative evaluation of numerous highly disparate yet coupled phenomenologies. These include two-phase thermodynamics and thermal-hydraulics, aerosol physics, fission product phenomena, core-concrete interactions, the formation and combustion of flammable gases, and performance of engineered safety features. In the past, this complexity has meant that a complete containment analysis would require application of suites of separate computer codes each of which would treat only a narrower subset of these phenomena, e.g., a thermal-hydraulics code, an aerosol code, a core-concrete interaction code, etc. In this paper, we describe the development and some recent applications of the CONTAIN code, which offers an integrated treatment of the dominant containment phenomena and the interactions among them. We describe the results of a series of containment phenomenology studies, based upon realistic accident sequence analyses in actual plants. These calculations highlight various phenomenological effects that have potentially important implications for source term and/or containment loading issues, and which are difficult or impossible to treat using a less integrated code suite

  4. Development of Severe Accident Containment Analysis Package

    Energy Technology Data Exchange (ETDEWEB)

    Park, Chang-Hwan; Kim, Dong-Min; Seo, Jea-Uk; Lee, Dea-Young; Park, Soon-Ho; Lee, Jae-Gwon; Lee, Jin-Yong; Lee, Byung-Chul [FNC Technology Co., Yongin (Korea, Republic of)

    2016-10-15

    In safety viewpoint, the pressure and temperature of the containment is the important parameters, of course, the local hydrogen concentration is also the parameter of the major concern because of its flammability and the risk of the detonation. In addition, there are possibilities of occurrence of other relevant phenomena following the reactor core melting such as DCH(direct containment heating) due to HPME(high pressure melt ejection), steam explosion due to fuel-coolant interaction in the reactor cavity and molten core concrete interaction at the late stage. It is important to predict the containment responses during a severe accident by a reasonable accuracy for establishing of effective mitigation strategies and preparation of the safety features required. In this paper, the overview of the SACAP development status is presented. SACAP is developed so as to be able to analyze, so called, Ex-Vessel severe accident phenomena including thermal-hydraulics, combustible gas burn, direct containment heating, steam explosion and molten core-concrete interaction. At the parallel time, SACAP and In-Vessel analysis module named CSPACE are processed for integration through MPI communication coupling. Development of the integrated severe accident analysis code system will be completed in following one year to make the code revision zero to be released.

  5. Containment severe accident management - selected strategies

    International Nuclear Information System (INIS)

    Duco, J.; Royen, J.; Rohde, J.; Frid, W.; De Boeck, B.

    1994-01-01

    The OECD Nuclear Energy Agency (NEA) organized in June 1994, in collaboration with the Swedish Nuclear Power Inspectorate (SKI), a Specialist Meeting on Selected Containment Severe Accident Management Strategies, to discuss their feasibility, effectiveness, benefits and drawbacks, and long-term impact. The meeting focused on water reactors, mainly on existing systems. The technical content covered topics such as general aspects of accident management strategies in OECD Member countries, hydrogen management techniques and other containment accident management strategies, surveillance and protection of the containment function. The main conclusions of the meeting are summarized in the paper. (author)

  6. Severe accident mitigation through containment design

    International Nuclear Information System (INIS)

    Bergeron, K.D.

    1990-01-01

    Recent US Department of Energy plans to construct a Heavy Water Reactor for the production of defense nuclear materials have created a unique opportunity to explore ways to mitigate severe accident concerns in the design stage. Drawing on an extensive background in USNRC-sponsored severe accident work, Sandia National Laboratories has been exploring a number of Heavy Water New Production Reactor (HW-NPR) containment design strategies that might mitigate the consequences of a core-melt accident without greatly impacting construction cost or reactor operations. Severe accident specialists have undertaken these assessments with the intent of providing the plant designers with some of the phenomenological advantages and disadvantages of various mitigation strategies. This paper will highlight some of the more interesting concepts and summarize the results obtained. 9 refs., 2 tabs

  7. Severe accident mitigation through containment design

    International Nuclear Information System (INIS)

    Bergeron, K.D.

    1990-01-01

    Recent U.S. Department of Energy plans to construct a Heavy Water Reactor for the production of defense nuclear materials have created a unique opportunity to explore ways to mitigate severe accident concerns in the design stage. Drawing on an extensive background in US-NRC-sponsored severe accident work, Sandia National Laboratories has been exploring a number of Heavy Water New Production Reactor (HW-NPR) containment design strategies that might mitigate the consequences of a core-melt accident without greatly impacting construction cost or reactor operations. Severe accident specialists have undertaken these assessments with the intent of providing the plant designers with some of the phenomenological advantages and disadvantages of various mitigation strategies. This paper will highlight some of the more interesting concepts and summarize the results obtained. (author). 9 refs., 2 tabs

  8. Studies of severe accidents in light water reactors. Containment performance

    International Nuclear Information System (INIS)

    Hayns, M.R.; Phillips, D.W.; Young, R.L.D.

    1987-01-01

    The containment system of a LWR is an obvious component of the plant which performs an important safety function in preventing the release of fission products to the environment in the event of design basis accidents. With over 260 LWRs in service worldwide, and others still under construction, there is a considerable diversity of containment types and combinations of containment safeguards systems. All of these satisfy local regulatory requirements which are principally aimed at the design basis accidents, and these requirements naturally have a considerable uniformity. However, their design diversity becomes more relevant to the performance of the containment in severe accident conditions, and this aspect of containment performance is reviewed in this paper. The ability of the containment to mitigate severe accident consequences introduces the potential for accident management and recovery and this in turn points towards a range of new containment systems and concepts. PSA helps in judging these possibilities and in forming policies and procedures for accident management. It is perhaps in accident management that severe accident containment performance will be most beneficial in the future, and where additional effort in containment analysis will be focused

  9. Containment loading during severe core damage accidents

    International Nuclear Information System (INIS)

    Fermandjian, J.; Evrard, J.M.; Cenerino, C.; Berthion, Y.; Carvallo, G.

    1984-11-01

    The objective of the article is to study the influence of the state of the reactor cavity (dry or flooded) and of the corium coolability on the thermal-hydraulics in the containment in the case of an accident sequence involving core melting and subsequent containment basemat erosion, in a 900 MWe PWR unit. Calculations are performed by using the JERICHO thermal hydraulics code

  10. 40 CFR 191.13 - Containment requirements.

    Science.gov (United States)

    2010-07-01

    ... requirements. (a) Disposal systems for spent nuclear fuel or high-level or transuranic radioactive wastes shall... 40 Protection of Environment 24 2010-07-01 2010-07-01 false Containment requirements. 191.13 Section 191.13 Protection of Environment ENVIRONMENTAL PROTECTION AGENCY (CONTINUED) RADIATION PROTECTION...

  11. New developments in containment in-service inspection requirements

    International Nuclear Information System (INIS)

    Staffiera, J.E.

    1995-01-01

    Section 11 of the ASME Boiler and Pressure Vessel Code contains requirements for inservice inspection of nuclear power plant components. Development of ASME Code requirements for containment inservice inspection was begun in 1977, and in 1979 the first such requirements were published in the form of Code Case N-236. Formal inclusion of these requirements in Section 11 of the ASME Code occurred with publication of Subsection IWE, ''Rules for inservice Inspection of Class MC Components of Nuclear Power Plants,'' in the 1980 Edition, Winter 1981 Addenda. At that time, inspection emphasis on nuclear power construction and operation activities was placed on welds and welding processes associated with steel containments and metallic liners of concrete containments. The need for repair-welding requirements was necessitated by containment design modifications for conditions not considered in several original plant designs. Welds in steel containments and metallic liners of concrete containments have not required significant amounts of repair, however, degradation of base metal in containments has become a major concern. Various degradation mechanisms have been identified as potential causes of damage to containment surfaces, including fatigue, corrosion and material embrittlement due to long-term radiation exposure. As a result of these concerns, and in response to comments generated by the Committee to Review Generic Requirements (CRGR) of the NRC in its review of Subsection IWE, emphasis on weld-based inservice inspection was redirected toward a containment-surface inservice inspection program. Significant changes were made to accommodate this re-emphasis. The majority of these changes were published in the 1992 Edition, with the 1992 Addenda, of Subsection IWE. The NRC Proposed Rulemaking was issued for a 75-day public comment period in January, 1994. This period was extended at the request of nuclear industry organizations to allow for meaningful evaluation

  12. Research requirements for improved design of reinforced concrete containment structures

    International Nuclear Information System (INIS)

    Banerjee, A.K.; Holley, M.J. Jr.

    1978-01-01

    Reinforced concrete is a competitive material for the construction of nuclear power plant containment structures. However, the designer is constrained by limited data on the behavior of certain construction details which require him to use what may be excessive rebar quantities and lead to difficult and costly construction. This paper discusses several design situations where research is recommended to increase the designer's options, to facilitate construction, and to extend the applicability of reinforced concrete to such changing containment requirements as may be imposed by an evolving nuclear technology. (Auth.)

  13. Severe accidents and nuclear containment integrity (SANCY). SANCY summary report

    Energy Technology Data Exchange (ETDEWEB)

    Lindholm, I. [VTT Processes, Espoo (Finland)

    2004-07-01

    SANCY project investigates physical phenomena related to severe nuclear accidents with importance to Finnish nuclear power plants. Currently the major topics are the ex-vessel coolability issues, long-term severe accident management and containment leak tightness and adoption and development of new calculation tools considering also the needs of the future Olkiluoto 3 plant. SANCY employs both experimental and analytical methods. (orig.)

  14. Shipping container response to three severe railway accident scenarios

    International Nuclear Information System (INIS)

    Mok, G.C.; Fischer, L.E.; Murty, S.S.; Witte, M.C.

    1998-01-01

    The probability of damage and the potential resulting hazards are analyzed for a representative rail shipping container for three severe rail accident scenarios. The scenarios are: (1) the rupture of closure bolts and resulting opening of closure lid due to a severe impact, (2) the puncture of container by an impacting rail-car coupler, and (3) the yielding of container due to side impact on a rigid uneven surface. The analysis results indicate that scenario 2 is a physically unreasonable event while the probabilities of a significant loss of containment in scenarios 1 and 3 are extremely small. Before assessing the potential risk for the last two scenarios, the uncertainties in predicting complex phenomena for rare, high- consequence hazards needs to be addressed using a rigorous methodology

  15. The role of nuclear reactor containment in severe accidents

    International Nuclear Information System (INIS)

    1989-04-01

    The containment is a structural envelope which completely surrounds the nuclear reactor system and is designed to confine the radioactive releases in case of an accident. This report summarises the work of an NEA Senior Group of Experts who have studied the potential role of containment in accidents exceeding design specifications (so-called severe accidents). Some possibilities for enhancing the ability of plants to reduce the risk of significant off-site consequences by appropriate management of the acident have been examined

  16. Some outstanding issues in severe accidents containment performance

    International Nuclear Information System (INIS)

    Sehgal, B.R.

    2004-01-01

    This paper describes the current status of the outstanding issues in severe accident performance of Light Water Reactor containments that have been raised in the last several years. The results of the research that has been performed on the topics concerning these issues will be described. Some of these issues have been resolved, some are close to resolution, while others need further evaluation and research results. (author)

  17. Numerical Study of Severe Accidents on Containment Venting Conditions

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Na Rae; Bang, Young Suk; Park, Tong Kyu; Lee, Doo Yong [FNC Technology Co., Yongin (Korea, Republic of); Choi, Yu Jung; Lee, Sang Won; Kim, Hyeong Taek [KHNP-CRI, Daejeon (Korea, Republic of)

    2014-10-15

    Under severe accident, the containment integrity can be challenged due to over-pressurization by steam and non-condensable gas generation. According to Seismic Probabilistic Safety Assessment (PSA) result, the late containment failure by over-pressurization has been identified as the most probable containment failure mode. In addition, the analyses of Fukushima nuclear power plant accident reveal the necessity of the proper containment depressurization to prevent the large release of the radionuclide to environment. Containment venting has been considered as an effective approach to maintain the containment integrity from over-pressurization. Basic idea of containment venting is to relieve the pressure inside of the containment by establishing a flow path to the external environment. To ensure the containment integrity under over-pressure conditions, it is crucial to conduct the containment vent in a timely manner with a sufficient discharge flow rate. It is also important to optimize the vent line size to prevent additional risk of leakage and to install at the site with limited space availability. The purpose of this study is to identify the effective venting conditions for preventing the containment over-pressurization and investigate the vent flow characteristics to minimize the consequence of the containment ventilation.. In order that, thermodynamic behavior of the containment and the discharged flow depending on different vent strategies are analyzed and compared. The representative accident scenarios are identified by reviewing the Level 2 PSA result and the sensitivity analyses with varying conditions (i.e. vent line size and vent initiation pressure) are conducted. MAAP5 model for the OPR1000 Korea nuclear power plant has been used for severe accident simulations. Containment venting can be an effective strategy to prevent the significant failure of the containment due to over-pressurization. However, it should be carefully conducted because the vented

  18. Numerical Study of Severe Accidents on Containment Venting Conditions

    International Nuclear Information System (INIS)

    Lee, Na Rae; Bang, Young Suk; Park, Tong Kyu; Lee, Doo Yong; Choi, Yu Jung; Lee, Sang Won; Kim, Hyeong Taek

    2014-01-01

    Under severe accident, the containment integrity can be challenged due to over-pressurization by steam and non-condensable gas generation. According to Seismic Probabilistic Safety Assessment (PSA) result, the late containment failure by over-pressurization has been identified as the most probable containment failure mode. In addition, the analyses of Fukushima nuclear power plant accident reveal the necessity of the proper containment depressurization to prevent the large release of the radionuclide to environment. Containment venting has been considered as an effective approach to maintain the containment integrity from over-pressurization. Basic idea of containment venting is to relieve the pressure inside of the containment by establishing a flow path to the external environment. To ensure the containment integrity under over-pressure conditions, it is crucial to conduct the containment vent in a timely manner with a sufficient discharge flow rate. It is also important to optimize the vent line size to prevent additional risk of leakage and to install at the site with limited space availability. The purpose of this study is to identify the effective venting conditions for preventing the containment over-pressurization and investigate the vent flow characteristics to minimize the consequence of the containment ventilation.. In order that, thermodynamic behavior of the containment and the discharged flow depending on different vent strategies are analyzed and compared. The representative accident scenarios are identified by reviewing the Level 2 PSA result and the sensitivity analyses with varying conditions (i.e. vent line size and vent initiation pressure) are conducted. MAAP5 model for the OPR1000 Korea nuclear power plant has been used for severe accident simulations. Containment venting can be an effective strategy to prevent the significant failure of the containment due to over-pressurization. However, it should be carefully conducted because the vented

  19. 42 CFR 84.174 - Respirator containers; minimum requirements.

    Science.gov (United States)

    2010-10-01

    ... 42 Public Health 1 2010-10-01 2010-10-01 false Respirator containers; minimum requirements. 84.174... Air-Purifying Particulate Respirators § 84.174 Respirator containers; minimum requirements. (a) Except..., durable container bearing markings which show the applicant's name, the type of respirator it contains...

  20. 42 CFR 84.74 - Apparatus containers; minimum requirements.

    Science.gov (United States)

    2010-10-01

    ... 42 Public Health 1 2010-10-01 2010-10-01 false Apparatus containers; minimum requirements. 84.74...-Contained Breathing Apparatus § 84.74 Apparatus containers; minimum requirements. (a) Apparatus may be equipped with a substantial, durable container bearing markings which show the applicant's name, the type...

  1. 30 CFR 47.41 - Requirement for container labels.

    Science.gov (United States)

    2010-07-01

    ... 30 Mineral Resources 1 2010-07-01 2010-07-01 false Requirement for container labels. 47.41 Section... TRAINING HAZARD COMMUNICATION (HazCom) Container Labels and Other Forms of Warning § 47.41 Requirement for container labels. (a) The operator must ensure that each container of a hazardous chemical has a label. If a...

  2. Aerosol behavior in the reactor containment building during severe accident

    International Nuclear Information System (INIS)

    Berthion, Y.; Lhiaubet, G.; Gauvain, J.

    1984-07-01

    Thermohydraulic behavior inside a PWR containment during severe accident depends on decay heat transferred to the sump water by aerosol gravitational settling and deposition. Conversely, aerosol behavior depends on thermal hydraulic conditions, especially atmosphere moisture for soluble aerosol GsI, and CsOH. Therefore, a small iterative procedure between thermo-hydraulic and aerosol calculations has been performed in order to evaluate the importance of this coupling between the two phenomena. In this paper, it is shown that with this procedure and using our codes JERICHO, RICOCHET and AEROSOLS/B1, the steam condensation on aerosols is an important phenomenon for a correct estimation of the attenuation factor of the suspended mass of aerosols in the airborne of the containment. Then, we have a more realistic assessment of the source term released by the containment

  3. Testing, verification and application of CONTAIN for severe accident analysis of LMFBR-containments

    International Nuclear Information System (INIS)

    Langhans, J.

    1991-01-01

    Severe accident analysis for LMFBR-containments has to consider various phenomena influencing the development of containment loads as pressure and temperatures as well as generation, transport, depletion and release of aerosols and radioactive materials. As most of the different phenomena are linked together their feedback has to be taken into account within the calculation of severe accident consequences. Otherwise no best-estimate results can be assured. Under the sponsorship of the German BMFT the US code CONTAIN is being developed, verified and applied in GRS for future fast breeder reactor concepts. In the first step of verification, the basic calculation models of a containment code have been proven: (i) flow calculation for different flow situations, (ii) heat transfer from and to structures, (iii) coolant evaporation, boiling and condensation, (iv) material properties. In the second step the proof of the interaction of coupled phenomena has been checked. The calculation of integrated containment experiments relating natural convection flow, structure heating and coolant condensation as well as parallel calculation of results obtained with an other code give detailed information on the applicability of CONTAIN. The actual verification status allows the following conclusion: a caucious analyst experienced in containment accident modelling using the proven parts of CONTAIN will obtain results which have the same accuracy as other well optimized and detailed lumped parameter containment codes can achieve. Further code development, additional verification and international exchange of experience and results will assure an adequate code for the application in safety analyses for LMFBRs. (orig.)

  4. Containment pressure monitoring method after severe accident in nuclear power plant

    International Nuclear Information System (INIS)

    Luo Chuanjie; Zhang Shishui

    2011-01-01

    The containment atmosphere monitoring system in nuclear power plant was designed on the basis of design accident. But containment pressure will increase greatly in a severe accident, and pressure instrument in the containment can't satisfy the monitoring requirement. A new method to monitor the pressure change in the containment after a severe accident was considered, through which accident soften methods can be adopted. Under present technical condition, adding a pressure monitoring channel out of containment for post-severe accident is a considerable method. Daya Bay Nuclear Power Plant implemented this modification, by which the containment release time can be delayed during severe accident, and nuclear safety can be increased. After analysis, this method is safe and feasible. (authors)

  5. 42 CFR 84.134 - Respirator containers; minimum requirements.

    Science.gov (United States)

    2010-10-01

    ... 42 Public Health 1 2010-10-01 2010-10-01 false Respirator containers; minimum requirements. 84.134... Respirators § 84.134 Respirator containers; minimum requirements. Supplied-air respirators shall be equipped with a substantial, durable container bearing markings which show the applicant's name, the type and...

  6. 40 CFR 265.315 - Special requirements for containers.

    Science.gov (United States)

    2010-07-01

    ... 40 Protection of Environment 25 2010-07-01 2010-07-01 false Special requirements for containers..., STORAGE, AND DISPOSAL FACILITIES Landfills § 265.315 Special requirements for containers. Unless they are very small, such as an ampule, containers must be either: (a) At least 90 percent full when placed in...

  7. 42 CFR 84.1134 - Respirator containers; minimum requirements.

    Science.gov (United States)

    2010-10-01

    ... 42 Public Health 1 2010-10-01 2010-10-01 false Respirator containers; minimum requirements. 84... Combination Gas Masks § 84.1134 Respirator containers; minimum requirements. (a) Except as provided in paragraph (b) of this section each respirator shall be equipped with a substantial, durable container...

  8. 40 CFR 264.315 - Special requirements for containers.

    Science.gov (United States)

    2010-07-01

    ... 40 Protection of Environment 25 2010-07-01 2010-07-01 false Special requirements for containers... FACILITIES Landfills § 264.315 Special requirements for containers. Unless they are very small, such as an ampule, containers must be either: (a) At least 90 percent full when placed in the landfill; or (b...

  9. 42 CFR 84.197 - Respirator containers; minimum requirements.

    Science.gov (United States)

    2010-10-01

    ... 42 Public Health 1 2010-10-01 2010-10-01 false Respirator containers; minimum requirements. 84.197... Cartridge Respirators § 84.197 Respirator containers; minimum requirements. Respirators shall be equipped with a substantial, durable container bearing markings which show the applicant's name, the type and...

  10. 19 CFR 115.40 - Technical requirements for containers.

    Science.gov (United States)

    2010-04-01

    ... 19 Customs Duties 1 2010-04-01 2010-04-01 false Technical requirements for containers. 115.40...; DEPARTMENT OF THE TREASURY CARGO CONTAINER AND ROAD VEHICLE CERTIFICATION PURSUANT TO INTERNATIONAL CUSTOMS CONVENTIONS Procedures for Approval of Containers After Manufacture § 115.40 Technical requirements for...

  11. Containment performance of S-prism under severe BDB conditions

    International Nuclear Information System (INIS)

    Boardman, C.E.; Dubberley, A.E.; Hui, M.; Iwashige, K.

    2001-01-01

    S-PRISM is an advanced Fast Reactor plant design that utilizes compact modular pool-type reactors sized to enable factory fabrication and an affordable prototype test of a single Nuclear Steam Supply System (NSSS) for design certification at minimum cost and risk. Based on the success of the previous DOE sponsored Advanced Liquid Metal Reactor (ALMR) program GE has continued to develop and assess the technical viability and economic potential of an up-rated modular Fast Reactor called Super PRISM (S-PRISM). S-PRISM retains all of the key ALMR design features including passive reactor shutdown, passive shutdown heat removal, and passive reactor cavity cooling that were developed under an earlier DOE program. An additional feature of S-PRISM involves the use an innovative containment system that reduces the required design basis containment pressure by a factor of two through the use of a controlled venting system. The performance of this innovative containment system is evaluated and described in this paper. (author)

  12. 40 CFR 59.103 - Container labeling requirements.

    Science.gov (United States)

    2010-07-01

    ... 40 Protection of Environment 5 2010-07-01 2010-07-01 false Container labeling requirements. 59.103... National Volatile Organic Compound Emission Standards for Automobile Refinish Coatings § 59.103 Container... automobile refinish coating or coating component container or package, the day, month, and year on which the...

  13. 40 CFR 59.405 - Container labeling requirements.

    Science.gov (United States)

    2010-07-01

    ... 40 Protection of Environment 5 2010-07-01 2010-07-01 false Container labeling requirements. 59.405... National Volatile Organic Compound Emission Standards for Architectural Coatings § 59.405 Container... section on the coating container in which the coating is sold or distributed. (1) The date the coating was...

  14. 42 CFR 84.117 - Gas mask containers; minimum requirements.

    Science.gov (United States)

    2010-10-01

    ... 42 Public Health 1 2010-10-01 2010-10-01 false Gas mask containers; minimum requirements. 84.117... SAFETY AND HEALTH RESEARCH AND RELATED ACTIVITIES APPROVAL OF RESPIRATORY PROTECTIVE DEVICES Gas Masks § 84.117 Gas mask containers; minimum requirements. (a) Gas masks shall be equipped with a substantial...

  15. Numerical simulation of radioisotope's dependency on containment performance for large dry PWR containment under severe accidents

    International Nuclear Information System (INIS)

    Mehboob, Khurram; Xinrong, Cao; Ahmed, Raheel; Ali, Majid

    2013-01-01

    Highlights: • Calculation and comparison of activity of BURN-UP code with ORIGEN2 code. • Development of SASTC computer code. • Radioisotopes dependency on containment ESFs. • Mitigation in atmospheric release with ESFs operation. • Variation in radioisotopes source term with spray flow and pH value. -- Abstract: During the core melt accidents large amount of fission products can be released into the containment building. These fission products escape into the environment to contribute in accident source term. The mitigation in environmental release is demanded for such radiological consequences. Thus, countermeasures to source term, mitigations of release of radioactivity have been studied for 1000 MWe PWR reactor. The procedure of study is divided into five steps: (1) calculation and verification of core inventory, evaluated by BURN-UP code, (2) containment modeling based on radioactivity removal factors, (3) selection of potential accidents initiates the severe accident, (4) calculation of release of radioactivity, (5) study the dependency of release of radioactivity on containment engineering safety features (ESFs) inducing mitigation. Loss of coolant accident (LOCA), small break LOCA and flow blockage accidents (FBA) are selected as initiating accidents. The mitigation effect of ESFs on source term has been studied against ESFs performance. Parametric study of release of radioactivity has been carried out by modeling and simulating the containment parameters in MATLAB, which takes BURN-UP outcomes as input along with the probabilistic data. The dependency of iodine and aerosol source term on boric and caustic acid spray has been determined. The variation in source term mitigation with the variation of containment spray flow rate and pH values have been studied. The variation in containment retention factor (CRF) has also been studied with the ESF performance. A rapid decrease in source term is observed with the increase in pH value

  16. Requirements for containment system components in CANDU nuclear power plants

    International Nuclear Information System (INIS)

    1988-02-01

    This Standard specifies the requirements and establishes the rules for design, fabrication, and installation of pressure-retaining containment system components. In this Standard the term 'components' includes non registered items

  17. Requirements for containment system components in CANDU nuclear power plants

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1988-02-01

    This Standard specifies the requirements and establishes the rules for design, fabrication, and installation of pressure-retaining containment system components. In this Standard the term `components` includes non registered items.

  18. LWR and HTGR coolant dynamics: the containment of severe accidents

    International Nuclear Information System (INIS)

    Theofanous, T.G.; Gherson, P.; Nourbakhsh, H.P.; Hu, K.; Iyer, K.; Viskanta, R.; Lommers, L.

    1983-07-01

    This is the final report of a project containing three major tasks. Task I deals with the fundamental aspects of energetic fuel/coolant interactions (steam explosions) as they pertain to LWR core melt accidents. Task II deals with the applied aspects of LWR core melt accident sequences and mechanisms important to containment response, and includes consideration of energetic fuel/coolant interaction events, as well as non-explosive ones, corium material disposition and eventual coolability, and containment pressurization phenomena. Finally, Task III is concerned with HTGR loss of forced circulation accidents. This report is organized into three major parts corresponding to these three tasks respectively

  19. Post Fukushima requirement of containment filtered venting system in NPPS

    International Nuclear Information System (INIS)

    Deo, Anuj Kumar; Bera, S.; Nagrale, D.B.; Lakshmanan, S.P.; Baburajan, P.K.; Paul, U.K.; Gaikwad, A.J.

    2015-01-01

    Post Fukushima safety enhancement through provision of an additional layer of Defence-in-Depth in the existing and new Indian nuclear power plants has led to the need of containment filtered venting system (CFVS). The regulatory review of the design of CFVS is in progress. In order to assess the same, the regulatory knowledge base had to be generated on the current state of the art of the design of such a system by study of the international experience on this system available in the open literature. The regulatory stand on requirements and implementation status of the CFVS in various countries were also studied. The information available on design features of various kinds of venting systems, relevant design basis and/or acceptance criteria were collected for supporting the design safety review of the Indian CFVS under consideration. During the on-going regulatory review process several analyses have been carried out, some more are in progress, to support the deliberations and decision making. This paper presents the above mentioned information and the summary of the analyses carried out including the status and outcome. Important aspects of the design review and associated analyses are also presented in this paper which includes the descriptions of the work on CFD study of venturi atomization, thermal hydraulics studies, shielding analysis and source term estimation studies carried out by the regulatory body. (author)

  20. Thermal conditions and functional requirements for molten fuel containment

    International Nuclear Information System (INIS)

    Kang, C.S.; Torri, A.

    1980-05-01

    This paper discusses the configuration and functional requirements for the molten fuel containment system (MFCS) in the GCFR demonstration plant design. Meltdown conditions following a loss of shutdown cooling (LOSC) accident were studied to define the core debris volume for a realistic meltdown case. Materials and thicknesses of the molten fuel container were defined. Stainless steel was chosen as the sacrificial material and magnesium oxide was chosen as the crucible material. Thermal conditions for an expected quasi-steady state were analyzed. Highlights of the functional requirements which directly affect the MFCS design are discussed

  1. Several required OWL features for indigenous knowledge management systems

    CSIR Research Space (South Africa)

    Alberts, R

    2012-05-01

    Full Text Available This paper describes the features required of OWL (Web Ontology Language) to realise and enhance Indigenous Knowledge (IK) digital repositories. Several needs for Indigenous Knowledge management systems (IKMSs) are articulated, based on extensive...

  2. KAPP-3 and 4 containment pressure following postulated severe accident along with SAMG implementation

    International Nuclear Information System (INIS)

    Sharma, Sanjeev Kr.; Bhartia, D.K.; Mohan, Nalini; Malhotra, P.K.; Ghadge, S.G.; Chandra, Umesh

    2011-01-01

    Containment is an ultimate safety barrier which is designed to enclose whole reactor systems and to prevent the spread of active air-borne fission products. Studies are done to access its performance following severe accident i.e. Loss of Coolant Accident (LOCA) along with failure of Emergency Core Cooling System (ECCS), moderator and calandria vault water cooling system. The accident progression begins with the double ended break in reactor outlet/inlet header with simultaneous failure of ECCS followed by failure of moderator and calandria vault water cooling system. Initially decay heat and metal water reaction energy are assumed to be added to moderator water resulting in boiling of moderator and re-pressurization of containment due to steam addition. Subsequent to moderator boiling, decay heat and metal water reaction energy are assumed to be added to calandria vault water resulting in boiling and re-pressurization of containment due to steam addition. After moderator and calandria vault water have completely boiled off, rapid hydrogen generation would take place due to oxidation of pressure tubes and calandria tubes. In such accident scenario, the core is severely damaged. It will also lead to release of a large quantity of radio nuclides to containment atmosphere. To arrest the progression of accident, which can result in Severe Core damage and large amount of hydrogen production, which could leads to containment failure due to hydrogen deflagration or detonation, application of Severe Accident Management Guidelines (SAMG) has been studied. SAMG involve addition of water to calandria and calandria vault. It would result the boiling of the added water and consequent pressurization of containment. This paper presents the analysis for pressure-temperature of KAPP-3 and 4 containment following the postulated accident along with the application of Severe Accident Management Guidelines (SAMG). SAMG initiated action helps in arresting the progression of core

  3. Recent advances in severe accident technology - direct containment heating in advanced light water reactors

    International Nuclear Information System (INIS)

    Fontana, M.H.

    1993-01-01

    The issues affecting high-pressure melt ejection (HPME) and the consequential containment pressurization from direct containment heating (DCH), as they affect advanced light water reactors (ALWRs), specifically advanced pressurized water reactors (APWRs), were reviewed by the U.S. Department of Energy Advanced Reactor Severe Accident Program (ARSAP). Recommendations from ARSAP regarding the design of APWRs to minimize DCH are embodied within the Electric Power Research Institute ALWR Utility Requirements Document, which specifies (a) a large, strong containment; (b) an in-containment refueling water storage tank; (c) a reactor cavity configuration that minimizes energy transport to the containment atmosphere; and (d) a reactor coolant system depressurization system. Experimental and analytical efforts, which have focused on current-generation plants, and analyses for APWRs were reviewed. Although DCH is a subject of continuous research and considerable uncertainties remain, it is the judgment of the ARSAP that reactors complying with the recommended design requirements would have a low probability of early containment failure due to HPME and DCH

  4. Determination of optimal LWR containment design, excluding accidents more severe than Class 8

    International Nuclear Information System (INIS)

    Cave, L.; Min, T.K.

    1980-04-01

    Information is presented concerning the restrictive effect of existing NRC requirements; definition of possible targets for containment; possible containment systems for LWR; optimization of containment design for class 3 through class 8 accidents (PWR); estimated costs of some possible containment arrangements for PWR relative to the standard dry containment system; estimated costs of BWR containment

  5. A generic approach for steel containment vessel success criteria for severe accident loads

    International Nuclear Information System (INIS)

    Sammataro, R.F.; Solonick, W.R.; Edwards, N.W.

    1993-01-01

    Safety has been defined as the foremost design criterion for the Heavy Water New Production Reactor (NPR-HWR) by the U.S. DOE, Office of New Production Reactors (NP). The DOE-NP issued the Deterministic Severe Accident Criteria (DSAC) concept to guide the design of the NPR-HWR containment for resistance to severe accidents. The DSAC concept provides for a generic approach for containment vessel success criteria to predict the threshold of containment failure under severe accident loads. This concept consists of two parts: (1) Problem Statements and (2) Success Criteria. The paper is limited to a discussion of a success criteria. These criteria define acceptable containment response measures and limits for each problem statement. The criteria are based on the 'best estimate' of failure with no conservatism. Rather, conservatism, if required, is to be provided in the problem statements prepared by the designer and/or the regulatory authorities. The success criteria are presented on a multi-tiered basis for static pressure and temperature loadings, dynamic loadings, and missiles that may impact the containment. Within the static pressure and temperature loadings and the dynamic loadings, the criteria are separated into elastic analysis success criteria and inelastic analysis success criteria. Each of these areas, in turn, defines limits on either the stress or strain measures as well as on measures for buckling and displacements. The rationale upon which these criteria are based is contained in referenced documents. Rigorous validation of the criteria by comparison with results from analytical or experimental programs and application of the criteria to a containment design remain as future tasks. (orig./HP)

  6. Severe hypoglycaemia post-gastric bypass requiring partial pancreatectomy

    DEFF Research Database (Denmark)

    Patti, M E; McMahon, G; Mun, E C

    2005-01-01

    AIMS/HYPOTHESIS: Postprandial hypoglycaemia following gastric bypass for obesity is considered a late manifestation of the dumping syndrome and can usually be managed with dietary modification. We investigated three patients with severe postprandial hypoglycaemia and hyperinsulinaemia unresponsive...... was assessed in all three patients. RESULTS: All three patients had evidence of severe postprandial hyperinsulinaemia and hypoglycaemia. In one patient, reversal of gastric bypass was ineffective in reversing hypoglycaemia. All three patients ultimately required partial pancreatectomy for control...

  7. A Case of Severe and Recurrent Painless Thyroiditis Requiring Thyroidectomy

    Science.gov (United States)

    Ishii, Hiroaki; Takei, Masahiro; Sato, Yoshihiko; Ito, Tokiko; Ito, Ken-ichi; Sakai, Yasuhiro; Yumita, Wataru; Suzuki, Satoru; Komatsu, Mitsuhisa

    2013-01-01

    Objective To report a case of severe and recurrent painless thyroiditis requiring thyroidectomy. Clinical Presentation and Intervention A 47-year-old man who presented with severe thyrotoxicosis was found to have extremely low radioactive iodine uptake, negative TSH receptor antibodies, and normal C-reactive protein; these findings suggested a diagnosis of painless thyroiditis. Due to the severity and recurrence of thyrotoxicosis, surgical resection of the thyroid gland was performed to prevent a thyrotoxic storm. Histological examination revealed typical lymphoid infiltration of the thyroid gland. Conclusion This case illustrates that a patient with painless thyroiditis was successfully treated with surgery. PMID:23182952

  8. Estimate of the crud contribution to shipping cask containment requirements

    International Nuclear Information System (INIS)

    Sandoval, R.P.; Einziger, R.E.; Jordan, H.; Malinauskas, A.P.; Mings, W.J.

    1992-01-01

    This paper reports that a methodology is developed to relate U.S. Code of Federal Regulations, Title 10, Part 71 (10CFR71) containment requirements to leak rates for the special case in which the only radioactive species having a potential for escape form the cask is that associated with debris (crud) contained on the fuel assemblies being transported. The methodology accounts for the characteristics of the crud and for attenuation of the gas-borne crud particulates once they become suspended within the cask. Calculations are performed for typical spent-fuel transport cask geometries and the normal and accident conditions prescribed in 10CFR71. The most current published data are used for crud composition and structure, specific activity, spallation mechanics and fractions, and crud particle size. The containment criteria leak rates are calculated assuming 5-yr-old spent fuel. In each accident case, the containment leak rate criteria are well in excess of 10 cm 3 /s. Under normal conditions of transport, the regulatory containment requirements are met by leak rates ranging from 1.5 x 10 -3 cm 3 /s to 1.5 x 10 -4 cm 3 /s for the transport of boiling water reactor fuel assemblies and form 1.8 x 10 -2 cm 3 /s to 1.3 x 10 -3 cm 3 /s for pressurized water reactor fuel assemblies. The calculated leak rates are most sensitive to the cask design, type of fuel, and particle size distribution. Conservatism of the limiting leak rates is discussed

  9. Potential for containment leak paths through electrical penetration assemblies under severe accident conditions. [PWR; BWR

    Energy Technology Data Exchange (ETDEWEB)

    Sebrell, W.

    1983-07-01

    The leakage behavior of containments beyond design conditions and knowledge of failure modes is required for evaluation of mitigation strategies for severe accidents, risk studies, emergency preparedness planning, and siting. These studies are directed towards assessing the risk and consequences of severe accidents. An accident sequence analysis conducted on a Boiling Water Reactor (BWR), Mark I (MK I), indicated very high temperatures in the dry-well region, which is the location of the majority of electrical penetration assemblies. Because of the high temperatures, it was postulated in the ORNL study that the sealants would fail and all the electrical penetration assemblies would leak before structural failure would occur. Since other containments had similar electrical penetration assemblies, it was concluded that all containments would experience the same type of failure. The results of this study, however, show that this conclusion does not hold for PWRs because in the worst accident sequence, the long time containment gases stabilize to 350/sup 0/F. BWRs, on the other hand, do experience high dry-well temperatures and have a higher potential for leakage.

  10. Potential for containment leak paths through electrical penetration assemblies under severe accident conditions

    International Nuclear Information System (INIS)

    Sebrell, W.

    1983-07-01

    The leakage behavior of containments beyond design conditions and knowledge of failure modes is required for evaluation of mitigation strategies for severe accidents, risk studies, emergency preparedness planning, and siting. These studies are directed towards assessing the risk and consequences of severe accidents. An accident sequence analysis conducted on a Boiling Water Reactor (BWR), Mark I (MK I), indicated very high temperatures in the dry-well region, which is the location of the majority of electrical penetration assemblies. Because of the high temperatures, it was postulated in the ORNL study that the sealants would fail and all the electrical penetration assemblies would leak before structural failure would occur. Since other containments had similar electrical penetration assemblies, it was concluded that all containments would experience the same type of failure. The results of this study, however, show that this conclusion does not hold for PWRs because in the worst accident sequence, the long time containment gases stabilize to 350 0 F. BWRs, on the other hand, do experience high dry-well temperatures and have a higher potential for leakage

  11. Study of Containment Vent Strategies During Severe Accident Progression for the CANDU6 Plant

    Energy Technology Data Exchange (ETDEWEB)

    Jin, Youngho; Ahn, K. I. [KAERI, Daejeon (Korea, Republic of)

    2015-05-15

    In March, 2011, Fukushima daichi nuclear power plants experienced a long term station blackout. Severe core damage occurred and a large amount of radioactive materials are released outside of the plants. After this terrible accident Nuclear Safety and Security Commission (NSSC) enforced to increase nuclear safety for all operating plants in Korea. To increase plant safety, both hardware reinforcement and software improvement are encouraged. Hardware reinforcement includes the preparation of the external water injection paths to the RCS and the spent fuel pool, a filtered containment venting system (CFVS), and AC power generating truck. Software improvement includes the increase of the effectiveness of the severe accident management guidance (SAMG) and plant staff training. To comply with NSSC's request, Wolsong Unit 1 has fulfilled the hardware reinforcement including the installation of a CFVS and started the extension of a SAMG to the low power and shutdown operation mode. Current SAMG deals accident occurred during full power operation only. The CFVS is designed to open and to close isolation valves manually. It does not require AC power. The operation of the CFVS prevents the reactor containment building failure due to the over-pressurization but it may release radioactive materials out of the reactor containment building. This paper discusses the radiological source terms for the containment vent strategy during severe accident progression which occurred during shutdown operation mode. This work is a part of the development of shutdown SAMG.. The CFVS is an effective means to control the containment pressure when the local air coolers are unavailable. Radioactive materials may release through the CFVS, but their amounts are reduced significantly. The alternative means, i.e., containment vent through the ventilation system which does not have an effective filter, is not a good choice to control the containment condition. It can maintain the containment

  12. CE/Bechtel design containment response to severe accident phenomenology: A comparison among several combustion engineering plants

    International Nuclear Information System (INIS)

    Khalil, Y.F.; Schneider, R.E.

    1995-01-01

    The objectives of this paper are to: (1) discuss the types of severe accident phenomena that drive containment failure modes in CE plants and (2) contribute to the current state of knowledge of CE/Bechtel-design containment response to severe accident phenomenology. The second objective is addressed by providing a comparative study of containment response to severe accidents among several CE plants including Millstone Unit 2 (MP2), Palisades (Consumers Power), Calvert Cliffs (Baltimore Gas and Electric Company), Palo Verde (Arizona Public Service), and SONGS Units 2 and 3 (Southern California Edison). The motivation for addressing the second objective is based on the current lack of comprehensive literature on CE/Bechtel design containment failure modes and mechanisms for accidents that progress beyond the design basis limits. The first part of this paper addresses severe accident phenomena-related failure mechanisms in CE/Bechtel-designed containments. The second part of this work provides a comparative study of containment response among several CE plants

  13. 77 FR 23490 - Agency Information Collection Activities: Country of Origin Marking Requirements for Containers...

    Science.gov (United States)

    2012-04-19

    ... Activities: Country of Origin Marking Requirements for Containers or Holders AGENCY: U.S. Customs and Border... of Origin Marking Requirements for Containers or Holders. This is a proposed extension of an... Requirements for Containers or Holders. [[Page 23491

  14. Research and development strategy on the behavior of containments during severe accidents

    International Nuclear Information System (INIS)

    Lecomte, C.

    1990-06-01

    In case of an hypothetical severe accident leading to core melting, the last barrier preventing radionucleide release in the environnment is the containment of the main reactor building. The French research and development programmes aimed at understanding the containment behavior during severe accidents relate to several domains; some of them are: - assessment of hydrogen behavior - corium behavior and coolability - ultimate resistance of the containments and leaktightness - caracterization of filtered venting procedure. All these aspects are covered by code calculations and experimental developments

  15. KINS Research Activities on the iodine behavior in containment during a severe accident

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Hanchul; Kim, Dosam [Korea Institute of Nuclear Safety, Daejeon (Korea, Republic of); Oh, Jaeyong; Yun, Jongil [Korea Advanced Institute of Science and Technology, Daejeon (Korea, Republic of); Cho, Songwon [Korea Radiation Technology Institute, Daejeon (Korea, Republic of)

    2012-03-15

    Iodine is a major contributor to the potential health risk for the public following a severe accident from a nuclear power plant. Volatile iodine and organic iodides can be generated from the containment sump through various kinds of reactions and be released to the environment. This iodine behavior has been an important topic for the international research programs run by the OECD/NEA and EU-SARNET2. Korea Institute of Nuclear Safety (KINS) also has joined ISTP-EPICUR (Experimental Program on Iodine Chemistry under Radiation) and OECD-BIP (Behavior of Iodine Project). In the course of researching this issue with these experimental programs, a simple iodine model, RAIM, has been developed and coupled with the MELCOR code for radiological consequence analysis. This methodology is likely to provide a technical basis for developing the regulatory requirements concerning a severe accident including accident source term, which is one of urgent domestic needs.

  16. Modeling requirements for full-scope reactor simulators of fission-product transport during severe accidents

    International Nuclear Information System (INIS)

    Ellison, P.G.; Monson, P.R.; Mitchell, H.A.

    1990-01-01

    This paper describes in the needs and requirements to properly and efficiently model fission product transport on full scope reactor simulators. Current LWR simulators can be easily adapted to model severe accident phenomena and the transport of radionuclides. Once adapted these simulators can be used as a training tool during operator training exercises for training on severe accident guidelines, for training on containment venting procedures, or as training tool during site wide emergency training exercises

  17. Case report: Two severe cases of suicide attempts using nicotine containing e-cigarette liquid.

    Science.gov (United States)

    Jalkanen, Ville; Värelä, Ville; Kalliomäki, Jari

    The use of electronic cigarettes and nicotine-containing liquids is getting more common, thus increasing the risk for intentional or unintentional nicotine poisoning. The results of ingestion of nicotine can be severe, even fatal. We describe two different cases of severe poisonings caused by nicotine-containing electronic cigarette liquids.

  18. Specialist meeting on selected containment severe accident management strategies. Summary and conclusions

    International Nuclear Information System (INIS)

    1994-01-01

    The CSNI Specialist Meeting on Selected Containment Severe Accident Management Strategies held in Stockholm, Sweden in June 1994 was organised by the Task Group on Containment Aspects of Severe Accident Management (CAM) of CSNI's Principal Working Group on the Confinement of Accidental Radioactive Releases (PWG4) in collaboration with the Swedish Nuclear Power Inspectorate (SKI). Conclusions and recommendations are given for each of the sessions of the workshops: Containment accident management strategies (general aspects); hydrogen management techniques and other containment accident management techniques; surveillance and protection of containment function

  19. Phenomenological uncertainty analysis of containment building pressure load caused by severe accident sequences

    International Nuclear Information System (INIS)

    Park, S.Y.; Ahn, K.I.

    2014-01-01

    Highlights: • Phenomenological uncertainty analysis has been applied to level 2 PSA. • The methodology provides an alternative to simple deterministic analyses and sensitivity studies. • A realistic evaluation provides a more complete characterization of risks. • Uncertain parameters of MAAP code for the early containment failure were identified. - Abstract: This paper illustrates an application of a severe accident analysis code, MAAP, to the uncertainty evaluation of early containment failure scenarios employed in the containment event tree (CET) model of a reference plant. An uncertainty analysis of containment pressure behavior during severe accidents has been performed for an optimum assessment of an early containment failure model. The present application is mainly focused on determining an estimate of the containment building pressure load caused by severe accident sequences of a nuclear power plant. Key modeling parameters and phenomenological models employed for the present uncertainty analysis are closely related to the in-vessel hydrogen generation, direct containment heating, and gas combustion. The basic approach of this methodology is to (1) develop severe accident scenarios for which containment pressure loads should be performed based on a level 2 PSA, (2) identify severe accident phenomena relevant to an early containment failure, (3) identify the MAAP input parameters, sensitivity coefficients, and modeling options that describe or influence the early containment failure phenomena, (4) prescribe the likelihood descriptions of the potential range of these parameters, and (5) evaluate the code predictions using a number of random combinations of parameter inputs sampled from the likelihood distributions

  20. Proceedings of the specialist meeting on selected containment severe accident management strategies

    International Nuclear Information System (INIS)

    1995-07-01

    Twenty papers were presented at the first specialist meeting on Selected Containment Severe Accident management Strategies, held in Stockholm, Sweden, in 1994, half of them dealing with accident management strategies implementation status, half of them with research aspects. The four sessions were: general aspects of containment accident management strategies, hydrogen management techniques, other containment accident management strategies (spray cooling, core catcher...), surveillance and protection of containment function

  1. Proceedings of the specialist meeting on selected containment severe accident management strategies

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1995-07-15

    Twenty papers were presented at the first specialist meeting on Selected Containment Severe Accident management Strategies, held in Stockholm, Sweden, in 1994, half of them dealing with accident management strategies implementation status, half of them with research aspects. The four sessions were: general aspects of containment accident management strategies, hydrogen management techniques, other containment accident management strategies (spray cooling, core catcher...), surveillance and protection of containment function

  2. Role of BWR secondary containments in severe accident mitigation: issues and insights from recent analyses

    International Nuclear Information System (INIS)

    Greene, S.R.

    1988-01-01

    All commercial boiling water reactor (BWR) plants in the US employ primary containments of the pressure suppression design. These primary containments are surrounded and enclosed by a secondary containment consisting of a reactor building and refueling bay (MK I and MK II designs), a shield building, auxiliary building and fuel building (MK III), or an auxiliary building and enclosure building (Grand Gulf style MK III). Although secondary containment designs are highly plant specific, their purpose is to minimize the ground level release of radioactive material for a spectrum of traditional design basis accidents. While not designed for severe accident mitigation, these secondary containments might also reduce the radiological consequences of severe accidents. This issue is receiving increasing attention due to concerns that BWR MK I primary containment integrity would be lost should a significant mass of molten debris escape the reactor vessel during a severe accident. This paper presents a brief overview of domestic BWR secondary containment designs and highlights plant-specific features that could influence secondary containment severe accident survivability and accident mitigation effectiveness. Current issues surrounding secondary containment performance are discussed, and insights gained from recent ORNL secondary containment studies of Browns Ferry, Peach Bottom, and Shoreham are presented. Areas of significant uncertainty are identified and recommendations for future research are presented

  3. Role of BWR MK I secondary containments in severe accident mitigation

    International Nuclear Information System (INIS)

    Greene, S.R.

    1986-01-01

    The recent advent of detailed containment analysis codes such as CONTAIN and MELCOR has facilitated the development of the first large-scale, architectural-based BWR secondary containment models. During the past year ORNL has developed detailed, plant-specific models of the Browns Ferry and Peach Bottom secondary containments, and applied these models in a variety of studies designed to evaluate the role and effectiveness of BWR secondary containments in severe accident mitigation. The topology and basis for these models is discussed, together with some of the emerging insights from these studies

  4. Carbon monoxide - hydrogen combustion characteristics in severe accident containment conditions. Final report

    International Nuclear Information System (INIS)

    2000-03-01

    Carbon monoxide can be produced in severe accidents from interaction of ex-vessel molten core with concrete. Depending on the particular core-melt scenario, the type of concrete and geometric factors affecting the interaction, the quantities of carbon monoxide produced can vary widely, up to several volume percent in the containment. Carbon monoxide is a combustible gas. The carbon monoxide thus produced is in addition to the hydrogen produced by metal-water reactions and by radiolysis, and represents a possibly significant contribution to the combustible gas inventory in the containment. Assessment of possible accident loads to containment thus requires knowledge of the combustion properties of both CO and H 2 in the containment atmosphere. Extensive studies have been carried out and are still continuing in the nuclear industry to assess the threat of hydrogen in a severe reactor accident. However the contribution of carbon monoxide to the combustion threat has received less attention. Assessment of scenarios involving ex-vessel interactions require additional attention to the potential contribution of carbon monoxide to combustion loads in containment, as well as the effectiveness of mitigation measures designed for hydrogen to effectively deal with particular aspects of carbon monoxide. The topic of core-concrete interactions has been extensively studied; for more complete background on the issue and on the physical/thermal-hydraulics phenomena involved, the reader is referred to Proceedings of CSNI Specialists Meetings (Ritzman, 1987; Alsmeyer, 1992) and a State-of-Art Report (European Commission, 1995). The exact amount of carbon monoxide present in a reactor pit or in various compartments (or rooms) in a containment building is specific to the type of concrete and the accident scenario considered. Generally, concrete containing limestone and sand have a high percentage of CaCO 3 . Appendix A provides an example of results of estimates of CO and CO 2

  5. Shipping container response to severe highway and railway accident conditions: Appendices

    International Nuclear Information System (INIS)

    Fischer, L.E.; Chou, C.K.; Gerhard, M.A.; Kimura, C.Y.; Martin, R.W.; Mensing, R.W.; Mount, M.E.; Witte, M.C.

    1987-02-01

    Volume 2 contains the following appendices: Severe accident data; truck accident data; railroad accident data; highway survey data and bridge column properties; structural analysis; thermal analysis; probability estimation techniques; and benchmarking for computer codes used in impact analysis. (LN)

  6. Use of open source software in estimating the effects of a severe accident on the Mark II containment

    International Nuclear Information System (INIS)

    Sainz, E.; Arguelles, R.

    2015-09-01

    Because the spectrum of scenarios of severe accident before which must verify the integrity of the containment can be very broad, it arises here a calculation methodology to estimate the structural response of the containment without incurring in high costs for commercial software licenses, or in times and calculation excessive requirements. The capabilities of computer programs with license of open source, OpenFOAM for CFD calculations and Salome-Meca for thermal and mechanical calculations were tested. The methodology begins of the venting of mass and energy that are postulated inside the container and the values of the thermal and mechanical fields are obtained through the walls. (Author)

  7. Estimate of radionuclide release characteristics into containment under severe accident conditions

    International Nuclear Information System (INIS)

    Nourbakhsh, H.P.

    1993-11-01

    A detailed review of the available light water reactor source term information is presented as a technical basis for development of updated source terms into the containment under severe accident conditions. Simplified estimates of radionuclide release and transport characteristics are specified for each unique combination of the reactor coolant and containment system combinations. A quantitative uncertainty analysis in the release to the containment using NUREG-1150 methodology is also presented

  8. Numerical simulation of radioisotope's dependency on containment performance for large dry PWR containment under severe accidents

    Energy Technology Data Exchange (ETDEWEB)

    Mehboob, Khurram, E-mail: khurramhrbeu@gmail.com [College of Nuclear Science and Technology, Harbin Engineering University, 145-31 Nantong Street, Nangang District, Harbin, Heilongjiang 150001 (China); Xinrong, Cao [College of Nuclear Science and Technology, Harbin Engineering University, 145-31 Nantong Street, Nangang District, Harbin, Heilongjiang 150001 (China); Ahmed, Raheel [College of Automation, Harbin Engineering University, 145-31 Nantong Street, Nangang District, Harbin, Heilongjiang 150001 (China); Ali, Majid [College of Nuclear Science and Technology, Harbin Engineering University, 145-31 Nantong Street, Nangang District, Harbin, Heilongjiang 150001 (China)

    2013-09-15

    Highlights: • Calculation and comparison of activity of BURN-UP code with ORIGEN2 code. • Development of SASTC computer code. • Radioisotopes dependency on containment ESFs. • Mitigation in atmospheric release with ESFs operation. • Variation in radioisotopes source term with spray flow and pH value. -- Abstract: During the core melt accidents large amount of fission products can be released into the containment building. These fission products escape into the environment to contribute in accident source term. The mitigation in environmental release is demanded for such radiological consequences. Thus, countermeasures to source term, mitigations of release of radioactivity have been studied for 1000 MWe PWR reactor. The procedure of study is divided into five steps: (1) calculation and verification of core inventory, evaluated by BURN-UP code, (2) containment modeling based on radioactivity removal factors, (3) selection of potential accidents initiates the severe accident, (4) calculation of release of radioactivity, (5) study the dependency of release of radioactivity on containment engineering safety features (ESFs) inducing mitigation. Loss of coolant accident (LOCA), small break LOCA and flow blockage accidents (FBA) are selected as initiating accidents. The mitigation effect of ESFs on source term has been studied against ESFs performance. Parametric study of release of radioactivity has been carried out by modeling and simulating the containment parameters in MATLAB, which takes BURN-UP outcomes as input along with the probabilistic data. The dependency of iodine and aerosol source term on boric and caustic acid spray has been determined. The variation in source term mitigation with the variation of containment spray flow rate and pH values have been studied. The variation in containment retention factor (CRF) has also been studied with the ESF performance. A rapid decrease in source term is observed with the increase in pH value.

  9. 77 FR 6815 - Agency Information Collection Activities: Country of Origin Marking Requirements for Containers...

    Science.gov (United States)

    2012-02-09

    ... Activities: Country of Origin Marking Requirements for Containers or Holders AGENCY: U.S. Customs and Border... information collection requirement concerning Country of Origin Marking Requirements for Containers or Holders...: Title: Country of Origin Marking Requirements for Containers or Holders. OMB Number: 1651-0057. Form...

  10. Options for management of containment integrity during severe accident in Indian PHWR

    International Nuclear Information System (INIS)

    Sharma, Sanjeev Kr.; Bhartia, D.K.; Mohan, N.; Nair, Suma R.

    2015-01-01

    Severe accident progressions have the potential to raise the containment pressure beyond the design pressure of the structure. Although the load withstanding capability of the containment structure has been assessed to be substantially higher than the design pressure of the structure (typically 2 times of design pressure), it is possible that a few components of Containment System may degrade leading to excessive release of radioactive fission gases at ground level. Additionally, possible cracks in the concrete of the containment at high pressure may aggravate the release at ground level. Over and above, maintaining high containment pressure high for a longer period increases the ground level release due to leakage from the containment, which effect on dose might be high. For maintaining the Integrity of the Containment, containment pressure can be reduced by either energy management system such as removing the heat from the calandria vault (CVWC) water by using CV water heat exchanger intermittently or reliving the containment atmosphere either through Primary Containment Controlled Discharge (PCCD) or Containment Filtered Venting System (CFVS). Further, it is necessary that these provisions must be initiated below design pressure. This paper presents the analysis for the containment depressurization by using CVWC system restored, manual opening of (PCCD) line and operation of CFVS during the progressions of the accident

  11. The assessment of containment codes by experiments simulating severe accident scenarios

    International Nuclear Information System (INIS)

    Karwat, H.

    1992-01-01

    Hitherto, a generally applicable validation matrix for codes simulating the containment behaviour under severe accident conditions did not exist. Past code applications have shown that most problems may be traced back to inaccurate thermalhydraulic parameters governing gas- or aerosol-distribution events. A provisional code-validation matrix is proposed, based on a careful selection of containment experiments performed during recent years in relevant test facilities under various operating conditions. The matrix focuses on the thermalhydraulic aspects of the containment behaviour after severe accidents as a first important step. It may be supplemented in the future by additional suitable tests

  12. Study on confinement function of reactor containment during late phase severe accident

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2013-08-15

    During a severe accident reactor containment integrity is maintained by accident management. However, gas leakage from containment is inevitable after the severe accident. A large amount of hydrogen and rare gases are produced due to core damage or melting. These non-condensable gases cause the containment pressure much higher than atmospheric pressure even after residual heat removal system recovery especially for BWR with smaller containment volume. Besides, iodine confined in water pool is re-evaporated under radiation field. The present study consists of realistic evaluation of fission products source term inside containment, quantitative evaluation of iodine re-evaporation effect and the experimental study of hydrogen treatment in BWR using ammonia production method by catalyst. Activities in fiscal year 2012 are that modification of MELCOR fission product chemical model was done and verified by experimental data, and that effects of CsI on ammonia production rate for Ru catalyst were conducted. (author)

  13. Some aspects of the research and development programmes on the behaviour of containments during severe accidents

    International Nuclear Information System (INIS)

    Dufresne, J.

    1989-01-01

    The R and D programmes relating to the behaviour of containments during severe accidents cover several domains: .leaktightness of the containment: this programme concerns the mechanical resistance of the concretes and the cracking criteria, on the one hand, and the leak rate through the porosities or cracks, on the other; . gaseous releases inside the containment. In addition to the releases of steam and fission products from the primary circuit, the gaseous H 2 0 and C0 2 releases from the concrete must also be studied: firstly during the corium-concrete interaction, and secondly during the heating of the internal surface of the containment which can be raised to a high temperature on contact with the atmosphere, for example during hydrogen combustion; . the release of fission products during the corium-concrete interactions; . the behaviour of the fission products inside the containment, particularly as regards iodine

  14. Insights into the behavior of LWR steel containment buildings during severe accidents

    International Nuclear Information System (INIS)

    Clauss, D.B.; Horschel, D.S.; Blejwas, T.E.

    1987-01-01

    Investigations into the performance of steel containment subject to pressure and temperature greater than their design basis loads are discussed. The timing, mechanism, and location of a containment failure, i.e., release of radioactive materials, have an important impact on the consequences of a severe accident. We review the results of experiments on steel containment models pressurized to failure, on aged and unaged seals subjected to elevated temperature and pressure, and on electrical penetration assemblies tested for leakage. Based on the results, the important features and details of analytical methods that can be used to predict containment performance are identified. Finally, we speculate on the performance of steel containments in severe accident conditions. (orig.)

  15. Thermal-hydraulic and aerosol containment phenomena modelling in ASTEC severe accident computer code

    International Nuclear Information System (INIS)

    Kljenak, Ivo; Dapper, Maik; Dienstbier, Jiri; Herranz, Luis E.; Koch, Marco K.; Fontanet, Joan

    2010-01-01

    Transients in containment systems of different scales (Phebus.FP containment, KAEVER vessel, Battelle Model Containment, LACE vessel and VVER-1000 nuclear power plant containment) involving thermal-hydraulic phenomena and aerosol behaviour, were simulated with the computer integral code ASTEC. The results of the simulations in the first four facilities were compared with experimental results, whereas the results of the simulated accident in the VVER-1000 containment were compared to results obtained with the MELCOR code. The main purpose of the simulations was the validation of the CPA module of the ASTEC code. The calculated results support the applicability of the code for predicting in-containment thermal-hydraulic and aerosol phenomena during a severe accident in a nuclear power plant.

  16. Recent insights from severe accident research and implications for containment criteria for advanced LWRs

    International Nuclear Information System (INIS)

    Speis, T.P.; King, T.L.; Eltawila, F.

    1992-01-01

    The Severe Accident Research Program (SARP) was begun after the TMI-2 accident in March, 1979. The rule for dealing with the generation of large quantity of hydrogen in BWRs and Ice Condenser PWRs was promulgated by the Nuclear Regulatory Commission (NRC). The NRC issued severe Accident Policy Statement in 1985, and the revised SARP in 1989. In this paper, the current understanding of the more important phenomena and the associated mechanical and thermal loads to the containment is described, and the on-going works are summarized. The containment loadings in severe accidents are listed, and direct containment heating and the liner failure in BWR Mark I are added. A great deal of informations obtained on the early phase of melt progression are shown. The current understanding of the severe accident phenomena related to the containment and the on-going related research efforts are discussed more in detail. Fuel-coolant interaction including alpha-mode containment failure, direct containment heating, hydrogen deflagration and detonation, core-concrete interaction and debris coolability are described. (K.I.)

  17. Failure Mode Estimation of Wolsong Unit 1 Containment Building with respect to Severe Accident Condition

    International Nuclear Information System (INIS)

    Hahm, Dae Gi; Choi, In Kil

    2009-01-01

    The containment buildings in a nuclear power plant (NPP) are final barriers against the exposure of harmful radiation materials at severe accident condition. Since the accident at Three Mile Island nuclear plant in 1979, it has become necessary to evaluate the internal pressure capacity of the containment buildings for the assessment of the safety of nuclear power plants. According to this necessity, many researchers including Yonezawa et al. and Hu and Lin analyzed the ultimate capacity of prestressed concrete containments subjected to internal pressure which can be occurred at sever accident condition. Especially in Wolsong nuclear power plant, the Unit 1 containment structures were constructed in the late 1970 to early 1980, so that the end of its service life will be reached in near future. Since that the complete decommission and reconstruction of the NPP may cause a huge expenses, an extension of the service time can be a cost-effective alternative. To extend the service time of NPP, an overall safety evaluation of the containment building under severe accident condition should be performed. In this study, we assessed the pressure capacity of Wolsong Unit 1 containment building under severe accident, and estimated the responses at all of the probable critical areas. Based on those results, we found the significant failure modes of Wolsong Unit 1 containment building with respect to the severe accident condition. On the other hand, for the aged NPP, the degradation of their structural performance must also be explained in the procedure of the internal pressure capacity evaluation. Therefore, in this study, we performed a parametric study on the degradation effects and evaluated the internal pressure capacity of Wolsong Unit 1 containment building with considering aging and degradation effects

  18. Development of a parametric containment event tree model for a severe BWR accident

    Energy Technology Data Exchange (ETDEWEB)

    Okkonen, T [OTO-Consulting Ay, Helsinki (Finland)

    1995-04-01

    A containment event tree (CET) is built for analysis of severe accidents at the TVO boiling water reactor (BWR) units. Parametric models of severe accident progression and fission product behaviour are developed and integrated in order to construct a compact and self-contained Level 2 PSA model. The model can be easily updated to correspond to new research results. The analyses of the study are limited to severe accidents starting from full-power operation and leading to core melting, and are focused mainly on the use and effects of the dedicated severe accident management (SAM) systems. Severe accident progression from eight plant damage states (PDS), involving different pre-core-damage accident evolution, is examined, but the inclusion of their relative or absolute probabilities, by integration with Level 1, is deferred to integral safety assessments. (33 refs., 5 figs., 7 tabs.).

  19. Experiments to evaluate behavior of containment piping bellows under severe accident conditions

    International Nuclear Information System (INIS)

    Lambert, L.D.; Parks, M.B.

    1993-01-01

    Bellows are an integral part of the containment pressure boundary in nuclear power plants. They are used at piping penetrations to allow relative movement between piping and the containment wall. In a severe accident they may be subjected to high pressure and temperature, and a combination of axial and lateral deflections. A test program to determine the leak-tight capacity of containment penetration bellows is being conducted at Sandia National Laboratories, Albuquerque, New Mexico. Several different bellows geometries, representative of actual containment bellows, are being subjected to extreme deflections along with pressure and temperature loads. The bellows geometries and loading conditions are described along with the testing apparatus and procedures. A total of thirteen tests have been conducted. The tests showed that withstanding relatively large bellows are capable of deformations, up to, or near, the point of full compression before developing leakage. The test data is presented and discussed

  20. Discussion on several important safety requirements for the new nuclear power plant

    International Nuclear Information System (INIS)

    Yan Tianwen; Li Jigen; Zhang Lin; Feng Youcai; Jia Xiang; Li Wenhong

    2013-01-01

    Post the Fukushima nuclear accident, the Chinese government raised higher safety goals and safety requirements for the new nuclear power plant to be constructed. The paper expounded the important indicators of safety requirements and the aspects of safety modification that had been developed for the new NPPs. It also discussed and analyzed the main fields required by the new NPPs safety requirements in the safety goals, safety evaluation of sites, defenses of internal and external events, severe accident prevention and mitigation, design of reactor core, containment system and I and C system, and optimization of engineering measure, which gave some references to the design, construction and safety modifications of new NPPs in China. (authors)

  1. Review of the severe accident risk reduction program (SARRP) containment event trees

    International Nuclear Information System (INIS)

    1986-05-01

    A part of the Severe Accident Risk Reduction Program, researchers at Sandia National Laboratories have constructed a group of containment event trees to be used in the analysis of key accident sequences for light water reactors (LWR) during postulated severe accidents. The ultimate goal of the program is to provide to the NRC staff a current assessment of the risk from severe reactor accidents for a group of five light water reactors. This review specifically focuses on the development and construction of the containment event trees and the results for containment failure probability, modes and timing. The report first gives the background on the program, the review criteria, and a summary of the observations, findings and recommendations. secondly, the individual reviews of each committee member on the event trees is presented. Finally, a review is provided on the computer model used to construct and evaluate the event trees

  2. Instrumentation availability for a pressurized water reactor with a large dry containment during severe accidents

    International Nuclear Information System (INIS)

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

    1991-03-01

    In support of the US Nuclear Regulatory Commission (NRC) Accident Management Research Program, the availability of instruments to supply accident management information during a broad range of severe accidents is evaluated for a pressurized water reactor with a large dry containment. Results from this evaluation include the following: (a) identification of plant conditions that would impact instrument performance and information needs during severe accidents, (b) definition of envelopes of parameters that would be important in assessing the performance of plant instrumentation for a broad range of severe accident sequences, and (c) assessment of the availability of plant instrumentation during severe accidents. 16 refs., 3 figs., 4 tabs

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

    International Nuclear Information System (INIS)

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

    1992-02-01

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

  4. Pressure Load Analysis during Severe Accidents for the Evaluation of Late Containment Failure in OPR-1000

    Energy Technology Data Exchange (ETDEWEB)

    Park, S. Y.; Ahn, K. I. [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2014-05-15

    The MAAP code is a system level computer code capable of performing integral analyses of potential severe accident progressions in nuclear power plants, whose main purpose is to support a level 2 probabilistic safety assessment or severe accident management strategy developments. The code employs lots of user-options for supporting a sensitivity and uncertainty analysis. The present application is mainly focused on determining an estimate of the containment building pressure load caused by severe accident sequences. Key modeling parameters and phenomenological models employed for the present uncertainty analysis are closely related to in-vessel hydrogen generation, gas combustion in the containment, corium distribution in the containment after a reactor vessel failure, corium coolability in the reactor cavity, and molten-corium interaction with concrete. The phenomenology of severe accidents is extremely complex. In this paper, a sampling-based phenomenological uncertainty analysis was performed to statistically quantify uncertainties associated with the pressure load of a containment building for a late containment failure evaluation, based on the key modeling parameters employed in the MAAP code and random samples for those parameters. Phenomenological issues surrounding the late containment failure mode are highly complex. Included are the pressurization owing to steam generation in the cavity, molten corium-concrete interaction, late hydrogen burn in the containment, and the secondary heat removal availability. The methodology and calculation results can be applied for the optimum assessment of a late containment failure model. The accident sequences considered were a loss of coolant accidents and loss of offsite accidents expected in the OPR-1000 plant. As a result, uncertainties addressed in the pressure load of the containment building were quantified as a function of time. A realistic evaluation of the mean and variance estimates provides a more complete

  5. Phenomenological uncertainty analysis of early containment failure at severe accident of nuclear power plant

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Su Won

    2011-02-15

    The severe accident has inherently significant uncertainty due to wide range of conditions and performing experiments, validation and practical application are extremely difficult because of its high temperature and pressure. Although internal and external researches were put into practice, the reference used in Korean nuclear plants were foreign data of 1980s and safety analysis as the probabilistic safety assessment has not applied the newest methodology. Also, it is applied to containment pressure formed into point value as results of thermal hydraulic analysis to identify the probability of containment failure in level 2 PSA. In this paper, the uncertainty analysis methods for phenomena of severe accident influencing early containment failure were developed, the uncertainty analysis that apply Korean nuclear plants using the MELCOR code was performed and it is a point of view to present the distribution of containment pressure as a result of uncertainty analysis. Because early containment failure is important factor of Large Early Release Frequency(LERF) that is used as representative criteria of decision-making in nuclear power plants, it was selected in this paper among various modes of containment failure. Important phenomena of early containment failure at severe accident based on previous researches were comprehended and methodology of 7th steps to evaluate uncertainty was developed. The MELCOR input for analysis of the severe accident reflected natural circulation flow was developed and the accident scenario for station black out that was representative initial event of early containment failure was determined. By reviewing the internal model and correlation for MELCOR model relevant important phenomena of early containment failure, the uncertainty factors which could affect on the uncertainty were founded and the major factors were finally identified through the sensitivity analysis. In order to determine total number of MELCOR calculations which can

  6. Hydrogen behavior in a large-dry pressurized water reactor containment building during a severe accident

    International Nuclear Information System (INIS)

    Hsu Wensheng; Chen Hungpei; Hung Zhenyu; Lin Huichen

    2014-01-01

    Following severe accidents in nuclear power plants, large quantities of hydrogen may be generated after core degradation. If the hydrogen is transported from the reactor vessel into the containment building, an explosion might occur, which might threaten the integrity of the building; this can ultimately cause the release of radioactive materials. During the Fukushima Daiichi nuclear accident in 2011, the primary containment structures remained intact but contaminated fragments broke off the secondary containment structures, which disrupted mitigation activities and triggered subsequent explosions. Therefore, the ability to predict the behavior of hydrogen after severe accidents may facilitate the development of effective nuclear reactor accident management procedures. The present study investigated the behavior of hydrogen in a large-dry pressurized water reactor (PWR). The amount of hydrogen produced was calculated using the Modular Accident Analysis Program. The hydrogen transport behavior and the effect of the explosion on the PWR containment building were simulated using the Flame Acceleration Simulator. The simulation results showed that the average hydrogen volume fraction is approximately 7% in the containment building and that the average temperature is 330 K. The maximum predicted pressure load after ignition is 2.55 bar, which does not endanger the structural integrity of the containment building. The results of this investigation indicate that the hydrogen mitigation system should be arranged on both the upper and lower parts of the containment building to reduce the impact of an explosion. (author)

  7. Severe accident consequence mitigation by filtered containment venting at Canadian nuclear power plants

    International Nuclear Information System (INIS)

    Lebel, Luke S.; Morreale, Andrew C.; Korolevych, Volodymyr; Brown, Morgan J.; Gyepi-Garbrah, Sam

    2017-01-01

    Highlights: • Use of filtered containment venting during a severe accident assessed. • Severe accident simulations performed using MAAP-CANDU and ADDAM. • Flow capacity, initiation protocols, efficiency, mass and thermal loading evaluated. • Efficient, robust system drastically reduces accident consequences. - Abstract: Having the capability to use filtered containment venting during a severe nuclear accident can significantly reduce its overall consequences. This study employs the MAAP-CANDU severe accident analysis code and the ADDAM atmospheric dispersion code to study the progression of: an unmitigated station blackout accident at a generic pressurized heavy water reactor, the release of radioactive material into the environment, the subsequent dispersion of the fission products through the atmosphere and the subsequent consequences (evacuation radius). The goal is to evaluate the application of filtered venting as an accident mitigation technology. Several aspects of filtered containment venting system design, like flow capacity, initiation protocols, filter efficiency, mass loading, and thermal loading are considered. An efficient and robust filtered containment venting system can reduce the amount of radiological materials emitted during an accident by 25 times or more, and as a result considerably reduce the off-site consequences of an accident.

  8. Severe Rhabdomyolysis Associated with Staphylococcus aureus Acute Endocarditis Requiring Surgery.

    Science.gov (United States)

    Ravry, Céline; Fedou, Anne-Laure; Dubos, Maria; Denes, Éric; Etchecopar, Caroline; Barraud, Olivier; Vignon, Philippe; François, Bruno

    2015-12-01

    Rhabdomyolysis has multiple etiologies with unclear mechanisms; however, rhabdomyolysis caused by Staphylococcus aureus infection is rare. A case report of severe rhabdomyolysis in a patient who presented with endocarditis caused by methicillin-susceptible S. aureus and review of relevant literature. The patient had a history of cardiac surgery for tetralogy of Fallot. He was admitted to the hospital because of fever and digestive symptoms. Respiratory and hemodynamic status deteriorated rapidly, leading to admission to the intensive care unit (ICU) for mechanical ventilation and vasopressor support. Laboratory tests disclosed severe rhabdomyolysis with a serum concentration of creatine kinase that peaked at 49,068 IU/L; all blood cultures grew methicillin-susceptible S. aureus. Antibiotic therapy was amoxicillin-clavulanic acid, ciprofloxacin, and gentamicin initially and was changed subsequently to oxacillin, clindamycin, and gentamicin. Transesophageal echocardiography showed vegetation on the pulmonary valve, thus confirming the diagnosis of acute endocarditis. Viral testing and computed tomography (CT) scan ruled out any obvious alternative etiology for rhabdomyolysis. Bacterial analysis did not reveal any specificity of the staphylococcal strain. The patient improved with antibiotics and was discharged from the ICU on day 26. He underwent redux surgery for valve replacement on day 53. Staphylococcal endocarditis should be suspected in cases of severe unexplained rhabdomyolysis with acute infectious symptoms.

  9. Analyses of a steel containment vessel with an outer contact structure under severe internal overpressurization conditions

    International Nuclear Information System (INIS)

    Porter, V.L.

    1994-01-01

    Many Mark-I and Mark-II BWR plants are designed with a steel vessel as the primary containment. Typically, the steel containment vessel (SCV) is enclosed within a reinforced concrete shield building with only a small gap (74-90 mm) separating the two structures. This paper describes finite element analyses performed to evaluate the effects of contact and friction between a steel containment vessel and an outer contact structure when the containment vessel is subjected to large internal pressures. These computations were motivated by a joint program on containment integrity involving the Nuclear Power Engineering Corporation (NUPEC) of Japan, the US Nuclear Regulatory Commission (NRC), and Sandia National Laboratories for testing model containments. Under severe accident loading conditions, the steel containment vessel in a typical Mark-I or Mark-II plant may deform under internal pressurization such that it contacts the inner surface of a shield building wall. (Thermal expansion from increasing accident temperatures would also close the gap between the SCV and the shield building, but temperature effects are not considered in these analyses.) The amount and location of contact and the pressure at which it occurs all affect how the combined structure behaves. A preliminary finite element model has been developed to analyze a model of a typical steel containment vessel con-ling into contact with an outer structure. Both the steel containment vessel and the outer contact structure were modelled with axisymmetric shell finite elements. Of particular interest are the influence that the contact structure has on deformation and potential failure modes of the containment vessel. Furthermore, the coefficient of friction between the two structures was varied to study its effects on the behavior of the containment vessel and on the uplift loads transmitted to the contact structure. These analyses show that the material properties of an outer contact structure and the amount

  10. The COLIMA experiment on aerosol retention in containment leak paths under severe nuclear accidents

    Energy Technology Data Exchange (ETDEWEB)

    Parozzi, Flavio, E-mail: flavio.parozzi@rse-web.it [RSE, Power Generation Department, via Rubattino 54, I-20134 Milano (Italy); Caracciolo, Eduardo D.J., E-mail: eduardo.caracciolo@rse-web.it [RSE, Power Generation Department, via Rubattino 54, I-20134 Milano (Italy); Journeau, Christophe, E-mail: christophe.journeau@cea.fr [CEA Cadarache (France); Piluso, Pascal, E-mail: pascal.piluso@cea.fr [CEA Cadarache (France)

    2013-08-15

    Highlights: ► Experiment investigating aerosol retention within concrete containment cracks under nuclear severe accident conditions. ► Provided representative conditions of the aerosols suspended inside the containment of PWRs under a severe accident. ► Prototypical aerosol particles generated with a thermite reaction and transported through the crack sample reproducing surface characteristics, temperature, pressure drop and gas leakage. ► The results indicate the significant retention due to zig-zag path. -- Abstract: CEA and RSE managed an experimental research concerning the investigation of aerosol retention within concrete containment cracks under severe accident conditions. The main experiment was carried out in November 2008 with aerosol generated from the COLIMA facility and a sample of cracked concrete with defined geometric characteristics manufactured by RSE. The facility provided representative conditions of the aerosols suspended inside the containment of PWRs under a severe accident. Prototypical aerosol particles were generated with a thermite reaction and transported through the crack sample, where surface characteristics, temperature, pressure drop and gas leakage were properly reproduced. The paper describes the approach adopted for the preparation of the cracked concrete sample and the dimensioning of the experimental apparatus, the test procedure and the measured parameters. The preliminary results, obtained from this single test, are also discussed in the light of the present knowledge about aerosol phenomena and the theoretical analyses of particle behaviour with the crack path.

  11. Introduction to Large-sized Test Facility for validating Containment Integrity under Severe Accidents

    International Nuclear Information System (INIS)

    Na, Young Su; Hong, Seongwan; Hong, Seongho; Min, Beongtae

    2014-01-01

    An overall assessment of containment integrity can be conducted properly by examining the hydrogen behavior in the containment building. Under severe accidents, an amount of hydrogen gases can be generated by metal oxidation and corium-concrete interaction. Hydrogen behavior in the containment building strongly depends on complicated thermal hydraulic conditions with mixed gases and steam. The performance of a PAR can be directly affected by the thermal hydraulic conditions, steam contents, gas mixture behavior and aerosol characteristics, as well as the operation of other engineering safety systems such as a spray. The models in computer codes for a severe accident assessment can be validated based on the experiment results in a large-sized test facility. The Korea Atomic Energy Research Institute (KAERI) is now preparing a large-sized test facility to examine in detail the safety issues related with hydrogen including the performance of safety devices such as a PAR in various severe accident situations. This paper introduces the KAERI test facility for validating the containment integrity under severe accidents. To validate the containment integrity, a large-sized test facility is necessary for simulating complicated phenomena induced by an amount of steam and gases, especially hydrogen released into the containment building under severe accidents. A pressure vessel 9.5 m in height and 3.4 m in diameter was designed at the KAERI test facility for the validating containment integrity, which was based on the THAI test facility with the experimental safety and the reliable measurement systems certified for a long time. This large-sized pressure vessel operated in steam and iodine as a corrosive agent was made by stainless steel 316L because of corrosion resistance for a long operating time, and a vessel was installed in at KAERI in March 2014. In the future, the control systems for temperature and pressure in a vessel will be constructed, and the measurement system

  12. Integral thermal model of severe accident dynamics of NPP with containment

    International Nuclear Information System (INIS)

    Arutyunyan, R.V.; Bol'shov, L.A.; Vasil'ev, A.D.; Kamennov, G.P.

    1991-01-01

    An analytical model of the interaction of reactor core remains with concrete during severe accidents at nuclear power plants is considered. Time dependences of side and radial concrete melting are plotted. Time dependences of containment atmosphere temperature and pressure during a severe accident at nuclear power plants are investigated analytically and numerically. The sensitivity of the results to the coefficient values in the problem is studied within the range of their concertainty. The Kaverna-1 is described. The results of modelling a severe NPP accident which have been obtained using the Kaverna-1 package are presented

  13. A generic approach for containment success criteria under severe accident loads

    International Nuclear Information System (INIS)

    Sammataro, R.F.; Solonick, W.R.; Edwards, N.W.

    1992-01-01

    The U.S. Department of Energy (DOE), Office of New Production Reactors (NP), has identified safety as the foremost design criterion for the Heavy Water New Production Reactor (NPR-HWR). The DOE-NP has issued the Deterministic Severe Accident Criteria (DSACs) to guide the design of the NPR-HWR containment for resistance to severe accidents. The DSAC concept provides for a generic approach for success criteria to predict the threshold of containment failure under severe accident loads. This concept consists of two parts: (1) Problem Statements that are qualitative and quantitative bases for calculating associated loadings and containment response to those loadings, and (2) Success Criteria that specify acceptable containment response measures and limits for each problem statement. This paper is limited to a discussion of a generic approach for containment success criteria. The main elements of these success criteria are expressed in terms of elastic stresses and inelastic strains. Containment performance is based on the best estimate of failure as predicted by either stress or strain, buckling, displacements, or ability to withstand missile perforation. Since these limits are best estimates of failure, no conservatism exists in these success criteria. Rather, conservatism is to be provided in the problem statements, i.e., the quantified severe accident loads. These success criteria are presented on a multi-tiered basis for static pressure and temperature loadings, dynamic loadings, and missiles. Within the static pressure and temperature loadings and the dynamic loadings, the criteria are separated into elastic analysis success criteria and inelastic analysis success criteria. Each of these areas, in turn, defines limits on either the stress or strain measures as well as on measures for buckling and displacements

  14. JERICHO computer code: PWR containment response during severe accidents description and sensitivity analysis

    International Nuclear Information System (INIS)

    Fermandjian, J.; Evrard, J.M.

    1983-12-01

    The JERICHO code has been developed in order to study the thermodynamic behaviour inside the reactor containment building for the complete spectrum of accident sequences likely to occur in such a reactor, including models for the various mass and energy transfer phenomena, for water spray, for hydrogen and carbon monoxide flammability limits and combustion, as well as for containment venting. Sensitivity analyses have been performed on a severe accident sequence, (namely, small LOCA with failure of the emergency core cooling and containment spray systems), involving core melting and subsequent concrete containment basemat erosion. The effect of various models, such as mass and energy transfer to the structures, has been studied. The influence of the concrete composition, of the fission product deposition and of the thermal degradation of the reactor cavity concrete walls on long term thermodynamic behaviour has also been investigated

  15. Containment failure modes preliminary analysis for Atucha-I nuclear power plant during severe accidents

    International Nuclear Information System (INIS)

    Baron, J.; Caballero, C.; Zarate, S.M.

    1997-01-01

    The present work has the objective to analyze the containment behavior of the Atucha-I nuclear power plant during a severe accident, as part of a probabilistic safety assessment (PSA). Initially, a generic description of the containment failure modes considered in other PSAs is performed. Then, the possible containment failure modes for Atucha I are qualitatively analyzed, according to it design peculiarities. These failure modes involve some substantial differences from other PSAs, due to the particular design of Atucha I. Among others, it is studied the influence of: moderator/coolant separation, existence of cooling Zircaloy channels, existence of filling bodies inside the pressure vessel, reactor cavity geometry, on-line refueling mode, and existence of a double shell containment (steel and concrete) with an annular separation room. As a functions of the before mentioning analysis, a series of parameters to be taken into account is defined, on a preliminary basis, for definition of the plant damage states. (author) [es

  16. NAUAHYGROS - A code for calculating aerosol behavior in nuclear power plant containments following a severe accident

    Energy Technology Data Exchange (ETDEWEB)

    Sher, R. [Rudolph Sher Associates, Stanford, CA (United States); Li, J. [Polestar Applied Technology, Inc., Los Altos, CA (United States)

    1995-02-01

    NAUAHYGROS is a computer code to calculate the behavior of fission product and other aerosol particles in the containment of a nuclear reactor following a severe accident. It is an extension of the German code NAUA, which has been in widespread use for many years. Early versions of NAUA treated various aerosol phenomena in dry atmospheres, including aerosol agglomeration, diffusion (plateout), and settling processes. Later versions added treatments of steam condensation on particles in saturated or supersaturated containment atmospheres. The importance of these condensation effects on aerosol removal rates was demonstrated in large scale simulated containment tests. The additional features incorporated in NAUAHYGROS include principally a treatment of steam condensation on hygroscopic aerosols, which can grow as a result of steam condensation even in superheated atmospheres, and improved modelling of steam condensation on the walls of the containment. The code has been validated against the LACE experiments.

  17. Bayesian optimization analysis of containment-venting operation in a boiling water reactor severe accident

    International Nuclear Information System (INIS)

    Zheng, Xiaoyu; Ishikawa, Jun; Sugiyama, Tomoyuki; Maryyama, Yu

    2017-01-01

    Containment venting is one of several essential measures to protect the integrity of the final barrier of a nuclear reactor during severe accidents, by which the uncontrollable release of fission products can be avoided. The authors seek to develop an optimization approach to venting operations, from a simulation-based perspective, using an integrated severe accident code, THALES2/KICHE. The effectiveness of the containment-venting strategies needs to be verified via numerical simulations based on various settings of the venting conditions. The number of iterations, however, needs to be controlled to avoid cumbersome computational burden of integrated codes. Bayesian optimization is an efficient global optimization approach. By using a Gaussian process regression, a surrogate model of the “black-box” code is constructed. It can be updated simultaneously whenever new simulation results are acquired. With predictions via the surrogate model, upcoming locations of the most probable optimum can be revealed. The sampling procedure is adaptive. Compared with the case of pure random searches, the number of code queries is largely reduced for the optimum finding. One typical severe accident scenario of a boiling water reactor is chosen as an example. The research demonstrates the applicability of the Bayesian optimization approach to the design and establishment of containment-venting strategies during severe accidents

  18. Bayesian optimization analysis of containment-venting operation in a boiling water reactor severe accident

    Energy Technology Data Exchange (ETDEWEB)

    Zheng, Xiaoyu; Ishikawa, Jun; Sugiyama, Tomoyuki; Maryyama, Yu [Nuclear Safety Research Center, Japan Atomic Energy Agency, Ibaraki (Japan)

    2017-03-15

    Containment venting is one of several essential measures to protect the integrity of the final barrier of a nuclear reactor during severe accidents, by which the uncontrollable release of fission products can be avoided. The authors seek to develop an optimization approach to venting operations, from a simulation-based perspective, using an integrated severe accident code, THALES2/KICHE. The effectiveness of the containment-venting strategies needs to be verified via numerical simulations based on various settings of the venting conditions. The number of iterations, however, needs to be controlled to avoid cumbersome computational burden of integrated codes. Bayesian optimization is an efficient global optimization approach. By using a Gaussian process regression, a surrogate model of the “black-box” code is constructed. It can be updated simultaneously whenever new simulation results are acquired. With predictions via the surrogate model, upcoming locations of the most probable optimum can be revealed. The sampling procedure is adaptive. Compared with the case of pure random searches, the number of code queries is largely reduced for the optimum finding. One typical severe accident scenario of a boiling water reactor is chosen as an example. The research demonstrates the applicability of the Bayesian optimization approach to the design and establishment of containment-venting strategies during severe accidents.

  19. The effect of severe accident mitigation design on the containment performance for Korean ALWR

    International Nuclear Information System (INIS)

    Na, J. H.; Lee, J. S.; Lim, H. K.; Kim, J. K.

    2001-01-01

    The containment performance analysis for Korean ALWR standard design has been performed to confirm the safety goal and to identify the design features vulnerable to severe accidents for the on-going design. The results in terms of conditional containment failure probability show Korean ALWR design does not have any particular vulnerability given core damage sequences. It shows the conditional containment failure probability for pull power internal event is less than that of design goal. The late containment failure is much less than 4% for given core damages and that of containment bypass is about 2%. New design features of the Korean ALWR such as bydrogen mitigation system (IIMS), cavity flooding system (CFS), and emergency containment spray bakcup system (ECSBS), external reactor vessel cooling (ERVC), etc. are reflected in Korean ALWR design and is reviewed in this paper to give an insight for the design vulnerabilities and input to the development of accident management. These Korean ALWR specific design features showed the containment performance is significantly enhanced compared with the other PWR plants

  20. Assessment of severe accident prevention and mitigation features: PWR, large dry containment design

    International Nuclear Information System (INIS)

    Perkins, K.R.; Hsu, C.J.; Lehner, J.R.; Luckas, W.J.; Cho, N.; Fitzpatrick, R.G.; Pratt, W.T.; Eltawila, F.; Maly, J.A.

    1988-07-01

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

  1. Assessment of severe accident prevention and mitigation features: PWR, ice-condenser containment design

    International Nuclear Information System (INIS)

    Hsu, C.J.; Perkins, K.R.; Luckas, W.J.; Fitzpatrick, R.G.; Cho, N.; Lehner, J.R.; Pratt, W.T.; Eltawila, F.; Maly, J.A.

    1988-07-01

    Plant features and operator actions which have been found to be important in either preventing and mitigating severe accidents in PWRs with ice-condenser containments have been identified. Thus features and actions were developed from insights derived from reviews of risk assessments performed specifically for the Sequoyah plant and from assessments of other relevant studies. Accident sequences that dominate the core-damage frequency and those accident sequences that are of potentially high consequence were identified. Vulnerabilities of the ice-condenser containment to sever accident containment loads were also identified. In addition, those features of a PWR with an ice-condenser containment, which are important for preventing core damage and are available for mitigating fission-product release to the environment were identified. This report is issued to provide focus to an analyst examining an individual plant. The report calls attention to plant features and operator actions and provides a list of deterministic attributes for assessing those features and actions found to be helpful in reducing the overall risk for Sequoyah and other PWRs with ice-condenser containments. Thus, the guidance is offered as a resource in examining the subject plant to determine if the same, or similar, plant features and operator actions will be of value in reducing overall plant risk. This report is intended to serve solely as guidance. 14 tabs

  2. Preliminary thermal design of a pressurized water reactor containment for handling severe accident consequences

    International Nuclear Information System (INIS)

    Abdullah, A.M.; Karameldin, A.

    1998-01-01

    A one-dimensional mathematical model has been developed for a 4250 MW(th) Advanced Pressurized Water Reactor containment analysis following a severe accident. The cooling process of the composite containment-steel shell and concrete shield- is achievable by natural circulation of atmospheric air. However, for purpose of gettering higher degrees of safety margin, the present study undertakes two objectives: (1) Installment of a diesel engine-driven air blower to force air through the annular space between the steel shell and concrete shield. The engine can be remotely operated to be effective in case of station blackout. (ii) Fixing longitudinally plate fins on the circumference of the inside and outside containment steel shell. These fins increase the heat transfer areas and hence the rate of heat removal from the containment atmosphere. In view of its importance - from the safety viewpoint - the long term behaviour of the containment which is a quasi-steady state problem, is formulated through a system of coupled nonlinear algebraic equations which describe the thermal-hydraulic and thermodynamic behaviour of the double shell containment. The calculated results revealed the following: (i) the passively air cooled containment can remove maximum heat load of 11.5 MW without failure, (ii) the effect of finned surface in the air passage tends to decrease the containment pressure by 20 to 30%, depending on the heat load, (iii) the effect of condensing fins is negligible for the proposed fin dimensions and material. However, by reducing the fin width, increasing their thickness, doubling their number, and using a higher conductive metal than the steel, it is expected that the containment pressure can be further reduced by 10% or more, (iv) the fins' dimensions and their number must be optimized via maximizing the difference or the ratio between the heat removed and pressure drop to get maximum heat flow rate

  3. γ radiation level simulation and analysis with MCNP in EPR containment during severe accident

    International Nuclear Information System (INIS)

    Zeng Jun; Liu Shuhuan; Wang Yang; Zhai Liang

    2013-01-01

    The γ dosimetry model based on the EPR core structure, material composition and the designed shielding system was established. The γ-ray dose rate distributions in EPR containment under different conditions including normal operation state, loss-of-coolant accident and core melt severe accident were simulated with MCNP5, and the calculation results under normal operation state and severe accident were compared and analyzed respectively with that of the designed limit. The study results may provide some relative data reference for EPR core accident prediction and reactor accident emergency decision making. (authors)

  4. Radionuclide release calculations for selected severe accident scenarios. Volume 3. PWR, subatmospheric containment design

    International Nuclear Information System (INIS)

    Denning, R.S.; Gieseke, J.A.; Cybulskis, P.; Lee, K.W.; Jordan, H.; Curtis, L.A.; Kelly, R.F.; Kogan, V.; Schumacher, P.M.

    1986-07-01

    This report presents results of analyses of the enviromental releases of fission products (source terms) for severe accident scenarios in a pressurized water reactor with a subatmospheric containment design. The analyses were performed to support the Severe Accident Risk Reduction/Risk Rebaselining Program (SARRP) which is being undertaken for the US Nuclear Regulatory Commission by Sandia National Laboratories. In the SARRP program, risk estimates are being generated for a number of reference plant designs. the Surry plant has been used in this study as the reference plant for a subatmospheric design

  5. An evaluation of alternate containment concepts for severe accident sequences: Chapter 3

    International Nuclear Information System (INIS)

    Ashton, D.H.; Blazo, S.R.

    1983-01-01

    Over the past several years, numerous design concepts have been developed to enhance the ability of containments to withstand severe reactor accidents. As part of the AIF sponsored IDCOR program, a study has been completed to survey and evaluate these alternate containment design concepts. The study defines the minimum as well as optimum functional and design criteria which any such system must meet. Six concepts which satisfy these criteria are then evaluated based upon factors such as: risk reduction potential, cost, constructability and the potential detrimental effects. Based upon the results of these evaluations, a ranking of the design concepts is developed. The purpose of this paper is to present the results of the IDCOR sponsored study

  6. An analysis of containment venting as a severe accident mitigation strategy for the BWR Mark II containment

    International Nuclear Information System (INIS)

    Kelly, D.L.; Galyean, W.J.

    1990-01-01

    An evaluation of a BWR/4 reactor with a Mark-II containment has identified the effects of containment venting on core damage frequency and containment failure mode, and has performed a limited evaluation of the effects on the off-site consequences. The analysis was founded upon an existing probabilistic risk assessment (PRA) with the addition of a proposed filtered containment venting system, based on the Swedish Filtra system installed at the Barseback nuclear power station in southern Sweden. Three different containment venting strategies were examined for their effects on plant risk. These are discussed

  7. Severe accident progression perspectives for Mark I containments based on the IPE results

    International Nuclear Information System (INIS)

    Lin, C.C.; Lehner, J.R.; Pratt, W.T.; Drouin, M.

    1995-01-01

    Based on level 2 analyses in IPE (Individual Plant Examination) submittals accident progression, perspectives were obtained for all containment types. These perspectives consisted of insights on containment failure modes, releases therein, and factors responsible for the results. To illustrate the types of perspectives acquired on severe accident progresssion, insights obtained for (BWR) Mark I containments are discussed here. Mark I containments have high strength but small volumes and rely on pressure suppression pools to condense steam released from the reactor coolant system during an accident. Accidents causing structural failure of the drywell shortly after the core debris melts through the reactor vessel were found to be dominant contributors to risk. Importance of individual containment failure mechanisms depends on plant features and in some cases on modeling assumptions; however the following mechanisms were found important: drywell shell melt-through caused by direct contact with core debris and drywell failure caused by rapid pressure/temperature pulses at time of vessel melt-through. Drywell failure caused by gradual pressure/temperature buildup due to gases and steam released during core/concrete interactions is important in some IPEs. In other IPEs vent was an important contributor. However, accidents that bypass containment (eg interfacing systems LOCA)or involve containment isolation failure were not important contributors to the CDF in any of the IPEs for Mark I plants. These accidents are also not important to risk (even though they can involve large fission product release) because their frequencies of occurrence are so much lower than frequencies of early structural failure caused by other accidents that dominate the CDF

  8. Analysis of flammability in the attached buildings to containment under severe accident conditions

    Energy Technology Data Exchange (ETDEWEB)

    Rosa, J.C. de la, E-mail: juan-carlos.de-la-rosa-blul@ec.europa.eu [European Commission Joint Research Centre (Netherlands); Fornós, Joan, E-mail: jfornosh@anacnv.com [Asociación Nuclear Ascó-Vandellós (Spain)

    2016-11-15

    Highlights: • Analysis of flammability conditions in buildings outside containment. • Stepwise approach easily applicable for any kind of containment and attached buildings layout. • Detailed application for real plant conditions has been included. - Abstract: Right after the events unfolded in Fukushima Daiichi, the European Union countries agreed in subjecting Nuclear Power Plants to Stress Tests as developed by WENRA and ENSREG organizations. One of the results as implemented in many European countries derived from such tests consisted of mandatory technical instructions issued by nuclear regulatory bodies on the analysis of potential risk of flammable gases in attached buildings to containment. The current study addresses the key aspects of the analysis of flammable gases leaking to auxiliary buildings attached to Westinghouse large-dry PWR containment for the specific situation where mitigating systems to prevent flammable gases to grow up inside containment are available, and containment integrity is preserved – hence avoiding isolation system failure. It also provides a full practical exercise where lessons learned derived from the current study – hence limited to the imposed boundary conditions – are applied. The leakage of gas from the containment to the support buildings is based on separate calculations using the EPRI-owned Modular Accident Analysis Program, MAAP4.07. The FATE™ code (facility Flow, Aerosol, Thermal, and Explosion) was used to model the transport and distribution of leaked flammable gas (H{sub 2} and CO) in the penetration buildings. FATE models the significant mixing (dilution) which occurs as the released buoyant gas rises and entrains air. Also, FATE accounts for the condensation of steam on room surfaces, an effect which acts to concentrate flammable gas. The results of the analysis show that during a severe accident, flammable conditions are unlikely to occur in compartmentalized buildings such as the one used in the

  9. Analysis of flammability in the attached buildings to containment under severe accident conditions

    International Nuclear Information System (INIS)

    Rosa, J.C. de la; Fornós, Joan

    2016-01-01

    Highlights: • Analysis of flammability conditions in buildings outside containment. • Stepwise approach easily applicable for any kind of containment and attached buildings layout. • Detailed application for real plant conditions has been included. - Abstract: Right after the events unfolded in Fukushima Daiichi, the European Union countries agreed in subjecting Nuclear Power Plants to Stress Tests as developed by WENRA and ENSREG organizations. One of the results as implemented in many European countries derived from such tests consisted of mandatory technical instructions issued by nuclear regulatory bodies on the analysis of potential risk of flammable gases in attached buildings to containment. The current study addresses the key aspects of the analysis of flammable gases leaking to auxiliary buildings attached to Westinghouse large-dry PWR containment for the specific situation where mitigating systems to prevent flammable gases to grow up inside containment are available, and containment integrity is preserved – hence avoiding isolation system failure. It also provides a full practical exercise where lessons learned derived from the current study – hence limited to the imposed boundary conditions – are applied. The leakage of gas from the containment to the support buildings is based on separate calculations using the EPRI-owned Modular Accident Analysis Program, MAAP4.07. The FATE™ code (facility Flow, Aerosol, Thermal, and Explosion) was used to model the transport and distribution of leaked flammable gas (H_2 and CO) in the penetration buildings. FATE models the significant mixing (dilution) which occurs as the released buoyant gas rises and entrains air. Also, FATE accounts for the condensation of steam on room surfaces, an effect which acts to concentrate flammable gas. The results of the analysis show that during a severe accident, flammable conditions are unlikely to occur in compartmentalized buildings such as the one used in the

  10. Containment response to a severe accident (TMLB sequence) with and without mitigation strategies

    International Nuclear Information System (INIS)

    Passalacqua, R.

    2004-01-01

    A loss of SG feed-water (TMLB sequence) for a prototypic PWR 900 MWe with a multi-compartment configuration (with 11 and 16 cells nodalization) has been calculated by the author using the ASTEC code in the frame of the EVITA project (5th Framework Programme, FWP). A variety of hypothesis (e.g. activation of sprays and hydrogen recombiners) and possible consequences of these assumptions (cavity flooding, hydrogen combustion, etc.) have been made in order to evaluate the global reactor containment building response (pressure, aerosol/FP concentration, etc.). The need to dispose of severe accident management guidelines (SAMGs) is increasing. These guidelines are meant for nuclear plants' operators in order to allow them to apply mitigation strategies all along a severe accident, which, only in its initial phase, may last several days. The purpose of this paper is to outline the influence on the containment load of most common accident occurrences and operators actions, which is essential in establishing SAMGs. ASTEC (Accident Source Term Evaluation Code) is a computer code for the evaluation of the consequences of a postulated nuclear plant severe accident sequence. ASTEC is a computer tool currently under joint development by the Institut de Radioprotection et de Surete Nucleaire (IRSN), France, and Gesellschaft fuer Anlagen-und Reaktorsicherheit (GRS), Germany. The aim of the development is to create a fast running integral code package, reliable in all simulations of a severe accident, to be used for level-2 PSA analysis. It must be said that several recent developments have significantly improved the best-estimate models of ASTEC and a new version (ASTEC V1.0) has been released mid-2002. Nevertheless, the somehow obsolete ASTECv0.3 version here used, has given results very useful for the estimation of the global risk of a nuclear plant. Moreover, under the current 6th FWP (Sustainable Integration of EU Research on Severe Accident Phenomenology and Management), the

  11. Fission products distributions in Candu primary heat transport and Candu containment systems during a severe accident

    International Nuclear Information System (INIS)

    Constantin, Marin; Rizoiu, Andrei

    2005-01-01

    The paper is intended to analyse the distribution of the fission products (FPs) in CANDU Primary Heat Transport (PHT) and CANDU Containment Systems by using the ASTEC code (Accident Source Term Evaluation Code). The complexity of the data required by ASTEC and the complexity both of CANDU PHT and Containment System were strong motivations to begin with a simplified geometry in order to avoid the introducing of unmanageable errors at the level of input deck. Thus only 1/4 of the PHT circuit was simulated and a simplified FPs inventory, some simplifications in the feeders geometry and containment were used. The circuit consists of 95 horizontal fuel channels connected to 95 horizontal out-feeders, then through vertical feeders to the outlet-header (a big pipe that collects the water from feeders); the circuit continues from the outlet-header with a riser and then with the steam generator and a pump. After this pump, the circuit was broken; in this point the FPs are transferred to the containment. The containment model consists of 4 rooms connected between by 6 links. The data related to the nodes' definitions, temperatures and pressure conditions were chosen as possible as real data from CANDU NPP loss of coolant accident sequence. Temperature and pressure conditions in the time of the accident were calculated by the CATHENA code and the source term of FPs introduced into the PHT was estimated by the ORIGEN code. The FPs distribution in the nodes of the circuit and the FPs mass transfer per isotope and chemical species are obtained by using SOPHAEROS module of ASTEC code. The distributions into the containment are obtained by the CPA module of ASTEC code (thermalhydraulics calculations in the containment and FPs aerosol transport). The results consist of mass distributions in the nodes of the circuit and the transferred mass to the containment through the break for different species (FPs and chemical species) and mass distributions in the different parts and

  12. WWER-440/V-230 Confinement modernization to upgrade the critical safety function 'Containment integrity' in case of severe accident

    International Nuclear Information System (INIS)

    Sartmadjiev, A.

    1999-01-01

    In this lecture the WWER-440/V-230 confinement modernization to upgrade the critical safety function 'Containment integrity' in case of severe accident is presented. There are discussed: design limitations of the location system; consequence from these design limitations; a few confinement reconstruction concepts of this type of units worldwide; and purpose of the confinement reconstruction - to improve significantly the original design, ensuring (1) localization for all possible primary breaks and (2) limitation of the radiological consequences for the personnel, the population and the environment below the regulatory requirements

  13. Modeling of Spray System Operation under Hydrogen and Steam Emissions in NPP Containment during Severe Accident

    Directory of Open Access Journals (Sweden)

    Vadim E. Seleznev

    2011-01-01

    Full Text Available The paper describes one of the variants of mathematical models of a fluid dynamics process inside the containment, which occurs in the conditions of operation of spray systems in severe accidents at nuclear power plant. The source of emergency emissions in this case is the leak of the coolant or rupture at full cross-section of the main circulating pipeline in a reactor building. Leak or rupture characteristics define the localization and the temporal law of functioning of a source of emergency emission (or accrued operating of warmed up hydrogen and steam in the containment. Operation of this source at the course of analyzed accident models should be described by the assignment of the relevant Dirichlet boundary conditions. Functioning of the passive autocatalytic recombiners of hydrogen is described in the form of the complex Newton boundary conditions.

  14. Preservice and inservice requirements for containment structures in the United States - a status report

    International Nuclear Information System (INIS)

    Sammataro, R.F.

    1987-01-01

    Assuring the lifetime integrity of containment structures for nuclear power plants is becoming increasingly important as existing design criteria are reexamined, as new requirements for containment inspection and testing are formulated, and as today's operating nuclear plants are growing older. Section XI of the ASME Code contains separate rules for metal (Class MC) and concrete (Class CC) containments. Requirements for Class MC containments have been published in Subsection IWE, Requirements for Class MC Components of Light-Water Cooled Power Plants, of Section XI. Rules for Class CC containments are currently being developed and will be published in Subsection IWL, Requirements for Class CC Components of Light-Water Cooled Power Plants, of Section XI. First published in 1981, Subsection IWE has been adopted by a number of state jurisdictions in the United States and is presently being reviewed by the United States Nuclear Regulatory Commission. Federal regulations that will require mandatory compliance by nuclear plant owners are forthcoming. When implemented, the requirements in Subsection IWE and Subsection IWL will provide a reasonable and systematic basis for assuring the integrity of metal and concrete containment structures during their service lifetime. This paper presents an overview of the preservice and inservice requirements for containment structures in Section XI of the ASME Code with consideration of the practical factors that should accompany user compliance. (orig./GL)

  15. Structural Integrity Evaluation of Containment Vessel under Severe Accident for PGSFR

    International Nuclear Information System (INIS)

    Lee, Seong-Hyeon; Koo, Gyeong-Hoi; Kim, Sung-Kyun

    2016-01-01

    This paper provides structural integrity evaluation results of CV of the PGSFR(Prototype Gen-IV Sodium Fast Reactor) under severe accident through transient analysis. The evaluation was carried out according to ASME B and PV Code Sec. III-Subsection NH rule. Structural integrity of CV was evaluated through transient analysis of structure in case of severe accident. Stress evaluation results for selected evaluation sections satisfy design criteria of ASME B and PV Code Sec. III Subsection NH. The transient load condition of normal operation will considered in the future work. The purpose of RVCS is to maintain the integrity of concrete structure during normal power operation. Therefore RVCS should be designed to keep the temperature of concrete surface under design limit and to minimize heat loss through CV(Containment Vessel). And in case of severe accident, the integrity of reactor structure and concrete structure should be maintained. Therefore RVCS should be designed to satisfy ASME Level D service limits. When RVCS works with breakdown of DHRS after severe accident, the temperature change of inner and outer surface of CV over time can affect structural integrity of CV. To verify the structural integrity, it is necessary to perform transient analysis of CV structure under changing temperature over time

  16. Stepwise integral scaling method for severe accident analysis and its application to corium dispersion in direct containment heating

    International Nuclear Information System (INIS)

    Ishii, M.; Zhang, G.; No, H. C.; Eltwila, F.

    1994-01-01

    Accident sequences which lead to severe core damage and to possible radioactive fission products into the environment have a very low probability. However, the interest in this area increased significantly due to the occurrence of the small break loss-of-coolant accident at TMI-2 which led to partial core damage, and of the Chernobyl accident in the former USSR which led to extensive core disassembly and significant release of fission products over several countries. In particular, the latter accident raised the international concern over the potential consequences of severe accidents in nuclear reactor systems. One of the significant shortcomings in the analyses of severe accidents is the lack of well-established and reliable scaling criteria for various multiphase flow phenomena. However, the scaling criteria are essential to the severe accident, because the full scale tests are basically impossible to perform. They are required for (1) designing scaled down or simulation experiments, (2) evaluating data and extrapolating the data to prototypic conditions, and (3) developing correctly scaled physical models and correlations. In view of this, a new scaling method is developed for the analysis of severe accidents. Its approach is quite different from the conventional methods. In order to demonstrate its applicability, this new stepwise integral scaling method has been applied to the analysis of the corium dispersion problem in the direct containment heating. ((orig.))

  17. Thermal stress analysis of reactor containment building considering severe weather condition

    International Nuclear Information System (INIS)

    Lee, Yun; Kim, Yun-Yong; Hyun, Jung-Hwan; Kim, Do-Gyeum

    2014-01-01

    Highlights: • We examine that through-wall crack risk in cold weather is high. • It is predicted that cracking in concrete wall will not happen in hot region. • Cracking due to hydration heat can be controlled by appropriate curing condition. • Temperature differences between inner and outer face is relatively small in hot weather. - Abstract: Prediction of concrete cracking due to hydration heat in mass concrete such as reactor containment building (RCB) in nuclear power plant is a crucial issue in construction site. In this study, the numerical analysis for heat transfer and stress development is performed for the containment wall in RCB by considering the severe weather conditions. Finally, concrete cracking risk in hot and cold weather is discussed based on analysis results. In analyses considering severe weather conditions, it is found that the through-wall cracking risk in cold weather is high due to the abrupt temperature difference between inside concrete and the ambient air in cold region. In hot weather, temperature differences between inner and outer face is relatively small, and accordingly the relevant cracking risk is relatively low in contrast with cold weather

  18. Assessment of severe accident prevention and mitigation features: BWR, Mark II containment design

    International Nuclear Information System (INIS)

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

    1988-07-01

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

  19. LWR severe accident simulation: Iodine behaviour in FPT2 experiment and advances on containment iodine chemistry

    Energy Technology Data Exchange (ETDEWEB)

    Girault, N., E-mail: nathalie.girault@irsn.fr [Institut de Radioprotection et de Surete Nucleaire (IRSN), BP3 - 13115 St.-Paul-lez-Durance (France); Bosland, L. [Institut de Radioprotection et de Surete Nucleaire (IRSN), BP3 - 13115 St.-Paul-lez-Durance (France); Dickinson, S. [National Nuclear Laboratory, Harwell, Oxon OX11 0QT (United Kingdom); Funke, F. [AREVA NP Gmbh, PO Box 1109, 91001 Erlangen (Germany); Guentay, S. [Paul Scherrer Institut, 5232 Villigen PSI (Switzerland); Herranz, L.E. [Centro des Investigaciones Energeticas, MedioAmbiantales y Tecnologicas, av. Complutense 2, 28040 Madrid (Spain); Powers, D. [Sandia National Laboratories, New Mexico, PO Box 5800, Albuquerque, NM 87185 (United States)

    2012-02-15

    Highlights: Black-Right-Pointing-Pointer Short term gaseous iodine fraction can be produced either in primary circuit or on containment condensing surfaces. Black-Right-Pointing-Pointer Gaseous radiolytic reactions convert volatile iodine into non-volatile iodine oxide particulates. Black-Right-Pointing-Pointer Alkaline and evaporating sump decrease the iodine volatility in containment. Black-Right-Pointing-Pointer Release of volatile iodine from containment surfaces explained the long term stationary residual gaseous iodine concentration. - Abstract: The Phebus Fission Product (FP) Program studies key phenomena of severe accidents in water-cooled nuclear reactors. In the framework of the Phebus program, five in-pile experiments have been performed that cover fuel rod degradation and behaviour of fission products released via the coolant circuit into the containment vessel. The focus of this paper is on iodine behaviour during the Phebus FPT2 test. FPT2 used a 33 GWd/t uranium dioxide fuel enriched to 4.5%, re-irradiated in situ for 7 days to a burn-up of 130 MWd/t. This test was performed to study the impact of steam-poor conditions and boric acid on the fission product chemistry. For the containment vessel, more specifically, the objective was to study iodine chemistry in an alkaline sump under evaporating conditions. The iodine results of the Phebus FPT2 test confirmed many of the essential features of iodine behaviour in the containment vessel provided by the first two Phebus tests, FPT0 and FPT1. These are the existence of an early gaseous iodine fraction, the persistence of low gaseous iodine concentrations and the importance of the sump in suppressing the iodine partitioning from sump to atmosphere. The main new insights provided by the Phebus FPT2 test were the iodine desorption from stainless steel walls deposits and the role of the evaporating sump in further iodine depletion in the containment atmosphere. The current paper presents an interpretation of

  20. Inflammatory responses and side effects generated by several adjuvant-containing vaccines in turbot.

    Science.gov (United States)

    Noia, M; Domínguez, B; Leiro, J; Blanco-Méndez, J; Luzardo-Álvarez, A; Lamas, J

    2014-05-01

    Several of the adjuvants used in fish vaccines cause adhesions in internal organs when they are injected intraperitoneally. We describe the damage caused by vaccines containing different adjuvants in the turbot Scophthalmus maximus and show that internal adhesions can be greatly reduced by injecting the fish in a specific way. Injection of fish with the needle directed towards the anterior part of the peritoneal cavity induced formation of a single cell-vaccine mass (CVM) that became attached to the parietal peritoneum. However, injection of the fish with the needle pointing in the opposite direction generated many small CVM that became attached to the visceral and parietal peritoneum and in some cases caused internal adhesions. We describe the structural and cellular changes in the adjuvant-induced CVMs. The CVMs mainly comprised neutrophils and macrophages, although most of the former underwent apoptosis, which was particularly evident from day 3 post-injection. The apoptotic cells were phagocytosed by macrophages, which were the dominant cell type from the first days onwards. All of the vaccines induced angiogenesis in the area of contact between the CVM and the mesothelium. Vaccines containing oil-based adjuvants or microspheres induced the formation of granulomas in the CVM; however, no granulomas were observed in the CVM induced by vaccines containing aluminium hydroxide or Matrix-Q(®) as adjuvants. All of the vaccines induced strong migration of cells to the peritoneal cavity. Although some of these cells remained unattached in the peritoneal cavity, most of them formed part of the CVM. We also observed migration of the cells from the peritoneal cavity to lymphoid organs, indicating bidirectional traffic of cells between the inflamed areas and these organs. Copyright © 2014 Elsevier Ltd. All rights reserved.

  1. Analytical capability for predicting structural response of NPP concrete containments to severe loads

    International Nuclear Information System (INIS)

    Planas, J.; Guinea, G.; Trbojevic, V.M.; Marti, J.; Martinez, F.; Cortes, P.

    1989-12-01

    A survey has been conducted on the state-of-the-art of analytical techniques for predicting the structural response of concrete containment buildings under severe accident conditions. The validity of inelastic analysis is often limited by the inadequacy of the material models adopted. This is specially true in the case of materials which undergo localization phenomena in the course of the deformation process. Because of this, the Joint Research Centre at Ispra has given a high priority to the review of existing constitutive models for concrete. Such models must be able to describe concrete behaviour with and without steel reinforcement across the complete stress range, from initial elastic behaviour to and beyond the point of failure. For reinforced and prestressed concrete, segregated models (where concrete and steel are independently simulated) are preferred. A review of existing constitutive models for mass concrete has been conducted. The review focused on necessary features for describing the near-peak and post-peak stages of deformation. Special attention was dedicated to the localization of strains in tension and the post-peak softening behaviour. Existing models for representing the concrete steel bond were also reviewed. These models are still relatively simplistic and incorporate seldom a number of effects of considerable importance: sustained, dynamic and cyclic loading, environmental effects, etc. Finally, the computational procedures currently available for modelling problems involving the ultimate capacity of concrete containments have also been reviewed. This includes methodologies for modelling amongst other mass concrete, cracking procedures, bond behaviour, in existing computer codes

  2. Raim – A model for iodine behavior in containment under severe accident condition

    Directory of Open Access Journals (Sweden)

    Han-Chul Kim

    2015-12-01

    Full Text Available Following a severe accident in a nuclear power plant, iodine is a major contributor to the potential health risks for the public. Because the amount of iodine released largely depends on its volatility, iodine's behavior in containment has been extensively studied in international programs such as International Source Term Programme-Experimental Program on Iodine Chemistry under Radiation (EPICUR, Organization for Economic Co-operation and Development (OECD-Behaviour of Iodine Project, and OECD-Source Term Evaluation and Mitigation. Korea Institute of Nuclear Safety (KINS has joined these programs and is developing a simplified, stand-alone iodine chemistry model, RAIM (Radio-Active Iodine chemistry Model, based on the IMOD methodology and other previous studies. This model deals with chemical reactions associated with the formation and destruction of iodine species and surface reactions in the containment atmosphere and the sump in a simple manner. RAIM was applied to a simulation of four EPICUR tests and one Radioiodine Test Facility test, which were carried out in aqueous or gaseous phases. After analysis, the results show a trend of underestimation of organic and molecular iodine for the gas-phase experiments, the opposite of that for the aqueous-phase ones, whereas the total amount of volatile iodine species agrees well between the experiment and the analysis result.

  3. RAIM-A model for iodine behavior in containment under severe accident condition

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Han Chul; Cho, Yeong Hun [Korea Institute of Nuclear Safety, Daejeon (Korea, Republic of)

    2015-12-15

    Following a severe accident in a nuclear power plant, iodine is a major contributor to the potential health risks for the public. Because the amount of iodine released largely depends on its volatility, iodine's behavior in containment has been extensively studied in international programs such as International Source Term Programme-Experimental Program on Iodine Chemistry under Radiation (EPICUR), Organization for Economic Co-operation and Development (OECD)-Behaviour of Iodine Project, and OECD-Source Term Evaluation and Mitigation. Korea Institute of Nuclear Safety (KINS) has joined these programs and is developing a simplified, stand-alone iodine chemistry model, RAIM (Radio-Active Iodine chemistry Model), based on the IMOD methodology and other previous studies. This model deals with chemical reactions associated with the formation and destruction of iodine species and surface reactions in the containment atmosphere and the sump in a simple manner. RAIM was applied to a simulation of four EPICUR tests and one Radioiodine Test Facility test, which were carried out in aqueous or gaseous phases. After analysis, the results show a trend of underestimation of organic and molecular iodine for the gas-phase experiments, the opposite of that for the aqueous-phase ones, whereas the total amount of volatile iodine species agrees well between the experiment and the analysis result.

  4. General requirements for concrete containment structures for CANDU nuclear power plants

    International Nuclear Information System (INIS)

    1993-07-01

    This standard provides the general requirements used in the design, construction, testing, and commissioning of concrete containment structures for CANDU nuclear power plants designated as class containment and is directed to the owners, designers, manufacturers, fabricators, and constructors of the concrete components and parts

  5. Examination and testing requirements for concrete containment structures for CANDU nuclear power plants

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1993-07-01

    This Standard provides the examination and testing requirements that will apply to the work of any organization participating in the construction, installation, and fabrication of parts or components of concrete containment structures, or both, that are defined as class containment. 2 tabs.

  6. Examination and testing requirements for concrete containment structures for CANDU nuclear power plants

    International Nuclear Information System (INIS)

    1993-07-01

    This Standard provides the examination and testing requirements that will apply to the work of any organization participating in the construction, installation, and fabrication of parts or components of concrete containment structures, or both, that are defined as class containment. 2 tabs

  7. Development of instrumentation systems for severe accidents. 4. New accident tolerant in-containment pressure transducer for containment pressure monitoring system

    International Nuclear Information System (INIS)

    Oba, Masato; Teruya, Kuniyuki; Yoshitsugu, Makoto; Ikeuchi, Takeshi

    2015-01-01

    The accident at Tokyo Electric Power Company's Fukushima Dai-ichi Nuclear Power Plant (TF-1 accident) caused severe situations and resulted in a difficulty in measuring important parameters for monitoring plant conditions. Therefore, we have studied the TF-1 accident to select the important parameters that should be monitored at the severe accident and are developing the Severe Accident Instrumentations and Monitoring Systems that could measure the parameters in severe accident conditions. Mitsubishi Heavy Industries, LTD (MHI) developed a new accident tolerant containment pressure monitoring system and demonstrated that the monitoring system could endure extremely harsh environmental conditions that envelop severe accident environmental conditions inside a containment such as maximum operating temperature of up to 300degC and total integrated dose (TID) of 1 MGy gamma. The new containment pressure monitoring system comprises of a strain gage type pressure transducer and a mineral insulated (MI) cable with ceramic connectors, which are located in the containment, and a strain measuring amplifier located outside the containment. Less thermal and radiation degradation is achieved because of minimizing use of organic materials for in-containment equipment such as the transducer and connectors. Several tests were performed to demonstrate the performance and capability of the in-containment equipment under severe accident environmental conditions and the major steps in this testing were run in the following test sequences: (1) the baseline functional tests (e.g., repeatability, non-linearity, hysteresis, and so on) under normal conditions, (2) accident radiation testing, (3) seismic testing, and (4) steam/temperature test exposed to simulated severe accident environmental conditions. The test results demonstrate that the new pressure transducer can endure the simulated severe accident conditions. (author)

  8. Problem of corium melt coolability in passive protection systems against severe accidents in the containment

    Directory of Open Access Journals (Sweden)

    Ali Kalvand

    2018-05-01

    Full Text Available Paper is devoted to the development of the mathematical model and analysis of the problem of corium melt interaction with low-temperature melting blocks in the passive protection systems against severe accidents at the NPP, which is of high importance for substantiation of the nuclear power safety, for building and successful op-erating of passive protection systems. In the third-generation reactors passive protection systems against severe accidents at the NPP are mandatory, therefore this paper is of importance for the nuclear power safety. A few configurations for the cooling blocks’ distribution have been considered and an analysis of the blocks’ melting and corium’s cooling in the pool under reactor vessel have been done, which can serve more effective for further improvement of the safety current systems and for the development of new ones. The ways for solution of the problems and the methods for their successful elaboration were discussed. The developed mathematical models and the analysis performed in the paper might be helpful for the design of passive protection systems of the cori-um melt retention inside the containment after corium melt eruption from the broken reactor vessel.

  9. PARIS project: Radiolytic oxidation of molecular iodine in containment during a nuclear reactor severe accident

    Energy Technology Data Exchange (ETDEWEB)

    Bosland, L. [Institut de Radioprotection et de Surete Nucleaire, DPAM, SEMIC, LETR - CEN Cadarache, BP 3, 13115 Saint Paul Lez Durance (France)], E-mail: loic.bosland@irsn.fr; Funke, F. [AREVA NP GmbH, Technical Center, P.O. Box 1109, D-91001 Erlangen (Germany); Girault, N. [Institut de Radioprotection et de Surete Nucleaire, DPAM, SEMIC, LETR - CEN Cadarache, BP 3, 13115 Saint Paul Lez Durance (France); Langrock, G. [AREVA NP GmbH, Technical Center, P.O. Box 1109, D-91001 Erlangen (Germany)

    2008-12-15

    In case of a hypothetical severe accident in a nuclear LWR (light water reactor), the high radiation fields reached in the reactor containment building due to the release of fission products from the reactor core could induce air radiolysis. The air radiolysis products could, in turn, oxidise gaseous molecular iodine into aerosol-borne iodine-oxygen-nitrogen compounds. Thereby, this reaction involves a change of iodine speciation and a decrease of iodine volatility in the reactor containment atmosphere. Kinetic data were produced within the PARIS project on the air radiolysis products formation and destruction, and on their reaction with molecular iodine, with the objective of developing and validating existing kinetic models. The current paper includes the non-iodine tests of the PARIS project whose objective was to determine the rates of formation and destruction of air radiolysis products in the presence of both structural containment surfaces (decontamination coating ('paint') and stainless steel), aerosol particles such as silver rich particles (issued from the control rods) in boundary conditions representative for LWR or PHEBUS facility containments. It is found that the air radiolysis products concentration increases with dose and tend to approach saturation levels at doses higher than about 1 kGy. This behaviour is more evident in oxygen/steam atmospheres, producing ozone, than in air/30% (v/v) steam atmospheres, the latter favouring the model-predicted on-going production of nitrogen dioxide even at very high doses. No significant effect of temperature, dose rate and hydrogen addition (4%, v/v) was observed. Furthermore, the inserted surfaces do not exhibit significant effects on the air radiolysis concentrations. However, these 'non-noticeable influence' could be due to a masking of small effects by the appreciable scattering of the experimental air radiolysis product concentrations. The PARIS results are then analysed using two

  10. PARIS project: Radiolytic oxidation of molecular iodine in containment during a nuclear reactor severe accident

    International Nuclear Information System (INIS)

    Bosland, L.; Funke, F.; Girault, N.; Langrock, G.

    2008-01-01

    In case of a hypothetical severe accident in a nuclear LWR (light water reactor), the high radiation fields reached in the reactor containment building due to the release of fission products from the reactor core could induce air radiolysis. The air radiolysis products could, in turn, oxidise gaseous molecular iodine into aerosol-borne iodine-oxygen-nitrogen compounds. Thereby, this reaction involves a change of iodine speciation and a decrease of iodine volatility in the reactor containment atmosphere. Kinetic data were produced within the PARIS project on the air radiolysis products formation and destruction, and on their reaction with molecular iodine, with the objective of developing and validating existing kinetic models. The current paper includes the non-iodine tests of the PARIS project whose objective was to determine the rates of formation and destruction of air radiolysis products in the presence of both structural containment surfaces (decontamination coating ('paint') and stainless steel), aerosol particles such as silver rich particles (issued from the control rods) in boundary conditions representative for LWR or PHEBUS facility containments. It is found that the air radiolysis products concentration increases with dose and tend to approach saturation levels at doses higher than about 1 kGy. This behaviour is more evident in oxygen/steam atmospheres, producing ozone, than in air/30% (v/v) steam atmospheres, the latter favouring the model-predicted on-going production of nitrogen dioxide even at very high doses. No significant effect of temperature, dose rate and hydrogen addition (4%, v/v) was observed. Furthermore, the inserted surfaces do not exhibit significant effects on the air radiolysis concentrations. However, these 'non-noticeable influence' could be due to a masking of small effects by the appreciable scattering of the experimental air radiolysis product concentrations. The PARIS results are then analysed using two different kinetic models

  11. Severe Accident Mitigation through Improvements in Filtered Containment Vent Systems and Containment Cooling Strategies for Water Cooled Reactors. Proceedings of a Technical Meeting

    International Nuclear Information System (INIS)

    2017-05-01

    One of the most important lessons from the accident at the Fukushima Daiichi nuclear power plant is that a reliable containment venting system can be crucial for effective accident management during severe accidents, especially for smaller volume containments in relation to the rated nuclear power. Containment venting can enhance the capability to maintain core cooling and containment integrity as well as reduce uncontrolled radioactive releases to the environment if the venting system has a filtration capacity. In general, a filtered containment vent system increases the flexibility of plant personnel in coping with unforeseen events. This publication provides the overview of the current status of related activities with the goal to share information between Member States on actions, upgrades, and new technologies pertaining to containment cooling and venting.

  12. Hyperammonemia following glufosinate-containing herbicide poisoning: a potential marker of severe neurotoxicity.

    Science.gov (United States)

    Mao, Yan-Chiao; Wang, Jiaan-Der; Hung, Dong-Zong; Deng, Jou-Fang; Yang, Chen-Chang

    2011-01-01

    Glufosinate-ammonium (GLA) is the active ingredient of certain widely used non-selective contact herbicides ("e.g.," Basta). Although it is thought to be much less toxic to humans than to plants, deliberate ingestion of GLA could still lead to serious effects ("e.g.," neurotoxicity) or even death. Three cases presented with delayed-onset neurotoxicity including stupor, delirium, seizures, coma, and amnesia after ingesting large amount of Basta. Considering that GLA could irreversibly inhibit glutamine synthetase (GS) in plants, we performed serial measurements of serum ammonia in those patients and revealed marked hyperammonemia in all of them. All patients recovered with the sequelae of persistent amnesia after receiving intensive care and hemodialysis. We speculated that the occurrence of hyperammonemia might at least be partially related to GS inhibition in humans. Moreover, hyperammonemia could serve as a potential marker of severe neurotoxicity, especially prolonged amnesia, following massive ingestion of GLA-containing herbicides. The possible dose-response relation between GLA exposure and serum ammonia level, however, needs more investigations.

  13. Nuclear waste shipping container response to severe accident conditions, A brief critique of the modal study

    International Nuclear Information System (INIS)

    Audin, L.

    1990-12-01

    The Modal Study (NUREG/CR-4829) attempts to upgrade the analysis of spent nuclear fuel transportation accidents, and to verify the validity of the present regulatory scheme of cask performance standards as a means to minimize risk. While an improvement over many prior efforts in this area (such as NUREG-0170), it unfortunately fails to create a realistic simulation either of a shipping cask, the severe conditions to which it could be subjected, or the potential damage to the spent fuel cargo during an accident. There are too many deficiencies in its analysis to allow acceptance of its results for the presumed cask design, and many pending changes in new containers, cargoes and shipping patterns will limit applicability of the Modal Study to future shipments. In essence, the Modal Study is a good start, but is too simplistic, incomplete, outdated and open to serious question to be used as the basis for any present-day environmental or risk assessment of spent fuel transportation. It needs to be redone, with peer review during its production and experimental verification of its assumptions, before it has any relevance to the shipments planned to Yucca Mountain. Finally, it must be expanded into a full risk assessment by inputing its radiological release fractions and probabilities into a valid dispersal simulation to properly determine the impact of its results. 51 refs

  14. Containment

    International Nuclear Information System (INIS)

    Anon.

    1987-01-01

    The primary mission of the Containment Group is to ensure that underground nuclear tests are satisfactorily contained. The main goal is the development of sound technical bases for containment-related methodology. Major areas of activity include siting, geologic description, emplacement hole stemming, and phenomenological predictions. Performance results of sanded gypsum concrete plugs on the Jefferson, Panamint, Cornucopia, Labquark, and Bodie events are given. Activities are also described in the following areas: computational capabilities site description, predictive modeling, and cavity-pressure measurement. Containment publications are listed. 8 references

  15. Impact of the filtered venting system design upon the total radioactive release in case of a severe accident and a comparison of European requirements

    International Nuclear Information System (INIS)

    Cederqvist, H.; Elisson, K.; Loewenhielm, G.; Appelgren, E.

    1991-01-01

    Filtered containment venting systems have been introduced in several nuclear power plants in Europe. The objective is to relieve the containment overpressure in a controlled way during a severe accident involving core-melt. The release of fission products when operating the venting system has been compared to that resulting from diffuse leakage from the containment. The conclusion is that the diffuse leakage of gaseous and particulate species can not be neglected in comparison to that resulting from operating the filtered containment venting system. Representative European requirements related to filtered containment venting have been analyzed and compared

  16. Containment hydrogen and atmosphere activity control to mitigate severe accidents in VVERs and Western PWRs. Design and status of implementation

    International Nuclear Information System (INIS)

    Feuerbach, R.

    2002-01-01

    For accident management nuclear power plants in Europe have been or will be back-fitted with supplementary systems for monitoring the containment hydrogen concentration, for the early removal and reduction of hydrogen and filtered venting systems to retain radioactive aerosols and iodine. The hydrogen monitoring system (HMS) provides the information of local H 2 concentration in the containment during DBA and severe accident situations. The new HMS contains of overall H 2 -sensors and is installed inside the confinement. It provides continuos information about the local and temporal distribution of hydrogen, reported directly to the Emergency Response Team in case of severe accident. The hydrogen Reduction System (HRS) consists of several Passive Autocatalytic Recombiners (PAR) located in several compartments in the containment. The number of PARs to be installed depends on the type of NPP, structure of containment and the investigated accident scenario e.g. DBA conditions - approx. 6 to 20 PARs; severe accident conditions - 20-60 PARs). In case of severe accident it does not need any operator actions. The Filtered Venting System (FVS) is is especially important for WWER-440/230 maintaining sub atmospheric pressure in the confinement. For severe accident the on-site Emergency Response Team has to take the necessary strategic decisions for containment depressurization via the FVS

  17. [Analysis of several containment measures of pharmaceutical expenditure in an Ambulatory Surgery Centre].

    Science.gov (United States)

    Esteban, J L; León, A; Porras, I

    2013-11-01

    In the context of the current crisis, sustainability of National Health Service must be considered a priority issue. To compare several cost saving measures in drug expenditure due to outpatient drug treatment after surgery in an Ambulatory Surgical Centre. Pharmaco-economic analysis of cost minimization of ambulatory pharmaceutical services during the year 2011. A total of 3,346 patients were operated on and discharged on the same day, were included. Treatments were collected from the discharge report of each patient. We compared changes in real outpatient drug spending after separately applying each of the following measures: 1) increasing the co-payment; 2) improving the quality of prescribing; 3) dispensing by units of drugs through pharmacies, and 4) dispensing through the hospital pharmacy service. The real outpatient pharmaceutical expenditure was 29,454.21€. Increasing the co-payment mean a transfer of 2,091.82€ from the funding institutions to users. Improving the quality of prescriptions, dispensing through units of drugs in the pharmacy, and dispensing through the hospital pharmacy service led to a pharmaceutical expenditure of 24,215.14€, 21,766.24€ and 7,827.71€, respectively. Only considering co-payment to contain pharmaceutical expenditure arising from prescribing in an Ambulatory Surgical Centre is the least effective measure. The most effective measure, for this purpose, is the supply of drugs through the hospital pharmacy service. Copyright © 2013 Sociedad Española de Anestesiología, Reanimación y Terapéutica del Dolor. Published by Elsevier España. All rights reserved.

  18. Analyses of a steel containment vessel with an outer contact structure under severe internal overpressurization conditions

    International Nuclear Information System (INIS)

    Porter, V.L.

    1993-01-01

    Many Mark-I and Mark-II BWR plants are designed with a steel vessel as the primary containment. Typically, the steel containment vessel (SCV) is enclosed within a reinforced concrete shield building with only a small gap (50--90mm) separating the two structures. This paper describes finite element analyses performed to evaluate the effects of contact and friction between a steel containment vessel and an outer contact structure when the containment vessel is subjected to large internal pressures. These computations were motivated by a joint program on containment integrity involving the Nuclear Power Engineering Corporation (NUPEC) of Japan, the US Nuclear Regulatory Commission (NRC), and Sandia National Laboratories for testing model containments

  19. A filter system for steam-gas mixture ejections from under a nuclear reactor containment following a severe accident

    International Nuclear Information System (INIS)

    Dulepov, Ju. N.; Sharygin, L. M.; Tretjakov, S. Ja.; Shtin, A.P.; Glushko, V. V.; Babenko, E. A.; Kurakov, Ju. A.

    1997-01-01

    In this paper newly built NPPs obligatory incorporate a containment having a filter system for removing radioactive materials ejections under severe accidents including nuclear fuel melting is described. The system prevents a containment failure and provides ejected radioactive materials decontamination to permissible levels. The physical-chemical and chemical characteristics of Termoxid-58 sorbent (TiO 5 based sorbent) are presented

  20. A study on the hydrogen distributions in a containment for nuclear plant severe accidents

    Energy Technology Data Exchange (ETDEWEB)

    Park, Kweon Ha; Kim, Ju Youn; Bae, Kyung Hyo [The Korea Maritime Univ., Busan (Korea, Republic of)

    2012-10-15

    Hydrogen explosion has been considered as one of the major issues since Fukushima nuclear accident. The cause of the explosion has not been discovered, but it is clear that the explosion strongly depends on hydrogen distributions in a containment. In this study hydrogen distributions are calculated and analyzed in the containment of APR 1400(Advanced Power Reactor 1400)

  1. Pre-operational proof and leakage rate testing requirements for concrete containment structures for CANDU nuclear power plants

    International Nuclear Information System (INIS)

    1994-02-01

    This Standard provides the requirements for pre-operational proof tests and leakage rate tests of concrete containment structures of a containment system designed as Class Containment components. 1 fig

  2. Tyrosine requirement during the rapid catch-up growth phase of recovery from severe childhood undernutrition

    Science.gov (United States)

    The requirement for aromatic amino acids, during the rapid catch-up in weight phase of recovery from severe childhood under nutrition (SCU) is not clearly established. As a first step, the present study aimed to estimate the tyrosine requirement of children with SCU during the catch-up growth phase ...

  3. A method for determining the spent-fuel contribution to transport cask containment requirements

    International Nuclear Information System (INIS)

    Sanders, T.L.; Seager, K.D.; Rashid, Y.R.; Barrett, P.R.; Malinauskas, A.P.; Einziger, R.E.; Jordan, H.; Reardon, P.C.

    1992-11-01

    This report examines containment requirements for spent-fuel transport containers that are transported under normal and hypothetical accident conditions. A methodology is described that estimates the probability of rod failure and the quantity of radioactive material released from breached rods. This methodology characterizes the dynamic environment of the cask and its contents and deterministically models the peak stresses that are induced in spent-fuel cladding by the mechanical and thermal dynamic environments. The peak stresses are evaluated in relation to probabilistic failure criteria for generated or preexisting ductile tearing and material fractures at cracks partially through the wall in fuel rods. Activity concentrations in the cask cavity are predicted from estimates of the fraction of gases, volatiles, and fuel fines that are released when the rod cladding is breached. Containment requirements based on the source term are calculated in terms of maximum permissible volumetric leak rates from the cask. Calculations are included for representative cask designs

  4. A study on the pressurized water reactor (PWR) containment response analysis methodologies for postulated severe accident

    International Nuclear Information System (INIS)

    Ahn, Kwang Il

    1992-02-01

    The present study contains two major parts: one is the treatment of uncertainties involved in the current APET and the other is the importance analysis of the APET uncertainty inputs. A clear disadvantage of the expert opinion polling process approach for uncertainty analysis of the current probabilistic risk assessment (PRA) is that the sufficient robustness in the final results may not be attained against the ambiguity of the information upon which the experts base their judgement or the judgmental uncertainty arising under various imprecise and incomplete information. For the treatment of such type of uncertainty, a new approach based on fuzzy set theory is proposed. Then its potential use to the uncertainty analysis of the current PRA is proved through an analysis of accident progression event tree (APET). As a product, a formal procedure with computational algorithms suitable for application of the fuzzy set theory to the APET analysis is provided. Comparing with the uncertainty analysis results obtained by the statistical approach currently used in PRA, the present approach has several major advantages: Firstly, it greatly enhances the robustness in the final results of APET uncertainty analysis by modeling the judgmental uncertainty that arises in the probabilistic quantification of APET top events. Secondly, the modeling of APET uncertainty analysis is far more convenient because of the nonprobabilistic features of fuzzy probabilities used for uncertainty quantification of the APET top events. Thirdly, the APET model can easily be operated by means of a well defined formal propagation logic of fuzzy set theory without going through a tedious sampling procedure. Finally, the fuzzy outcomes provide at least as much information as the existing methods based on the statistical approach. Thus, the present approach can be used as a valuable alternative approach to uncertainty analysis used in the current PRA. Two importance measures for the importance analysis of

  5. ORNL experiments to characterize fuel release from the reactor primary containment in severe LMFBR accidents

    International Nuclear Information System (INIS)

    Wright, A.L.; Kress, T.S.; Smith, A.M.

    1980-01-01

    This paper presents results from aerosol source term experiments performed in the ORNL Aerosol Release and Transport (ART) Program sponsored by the US NRC. The tests described were performed to provide information on fuel release from an LMFBR primary containment as a result of a hypothetical core-disruptive accident (HCDA). The release path investigated in these tests assumes that a fuel/sodium bubble is formed after disassembly that transports fuel and fission products through the sodium coolant and cover gas to be relased into the reactor secondary containment. Due to the excellent heat transfer characteristics of the sodium, there is potential for large attenuation of the maximum release

  6. French practice for assessing the fission product releases from the containment during a PWR severe accident

    International Nuclear Information System (INIS)

    Duco, J.; Dufresne, J.; L'homme, A.

    1988-10-01

    French safety philosophy as concerns severe PWR accidents has already been outlined by the Director of CEA/IPSN in an article published in ''Nuclear Safety''. Therefore the present paper will focus on: a) the French reference source terms, as used for elaborating ultimate emergency procedures on PWRs and for emergency planning; b) the methods currently developed for more realistic assessments of the release of fission products during a severe accident

  7. New system technologies implemented at Kozloduy 3 and 4 (WWER 440-230) for containment leakage and H2 control in severe accident situations - Design, qualification, installation, commissioning

    International Nuclear Information System (INIS)

    Feuerbach, R.; Eckardt, B.; Kastner, B.

    2005-01-01

    containment atmosphere and limited efficiency of retention, especially of the released airborne activities during overpressure periods through leaks. To mitigate the consequences of such postulated sever accident situations WWER 440-230 specific countermeasures have been developed, qualified and finally implemented at Kozloduy 3 and 4.The system technologies developed have to fulfill the following main tasks: 1) Creating a new pressure barrier system to confine the activity inventory by maintaining a slight containment sub-atmospheric pressure thereby preventing any relevant unfiltered activity releases; 2) Retaining a great extend of the airborne activity in the liquid phase of the venturi scrubber unit and return these activity back into the containment; 3) Controlling the containment atmosphere H 2 gas mixture to prevent fast deflagration that may jeopardize the containment integrity. To meet above mentioned requirement the system HIERARCH (High-Speed Venturi Activity Recirculation and Hold-up process) was developed, consisting of a venturi scrubber unit which comprises a high speed long venturi section, a metal fibre filter section and molecular sieve. The system is operated by a turbo blower unit supplied with electrical power from the unaffected unit. The system keeps a slight sub-atmospheric pressure (approx. 5 - 10 mbar) in the containment with the purpose to minimize the uncontrolled radioactive release from the containment to the environment. The system is designed for a throughput of up to 5 m/s steam gas mixture. The final analysis showed that the containment performance in combination with above mentioned countermeasures could be increased significantly and the impact to the environment is minimized. The radioactive doses beyond the border of the emergency planning area in case of severe accidents meet the Bulgarian requirements for BDBA

  8. Severe hypoglycaemia requiring the assistance of emergency medical services - frequency, causes and symptoms

    Czech Academy of Sciences Publication Activity Database

    Krnačová, V.; Kuběna, Aleš Antonín; Macek, K.; Bezděk, M.; Šmahelová, A.; Vlček, J.

    2012-01-01

    Roč. 156, č. 3 (2012), s. 271-277 ISSN 1213-8118 Grant - others:GA UK(CZ) SVV-2010-261-004 Keywords : regression trees * causes * symptoms * incidence * emergency medical service * severe hypoglycaemia Subject RIV: EI - Biotechnology ; Bionics Impact factor: 0.990, year: 2012 http://library.utia.cas.cz/separaty/2013/E/kubena-severe hypoglycaemia requiring the assistance of emergency medical services - frequency causes and symptoms.pdf

  9. Modeling of the corium cooling and loading factor analysis for containment during severe accidents

    International Nuclear Information System (INIS)

    Konoval, A.V.; Kalvand, Ali.; Kazachkov, I.V.

    2013-01-01

    The paper is devoted to the development and study of the mathematical model for corium melt interaction with low-temperature melting blocks in the passive protection systems (PPS) against severe accidents at the NPP, and learning the peculiarities of construction and operation of the PPS. The configurations of cooling blocks' distributions considered and the results of their work in the corium cooling pool are compared to the data of other PPS's conceptions. The conclusion is made that the models developed and the results obtained may be useful for constructing the PPS against severe accidents

  10. Modeling of the corium cooling and loading factor analysis for containment during severe accidents

    Directory of Open Access Journals (Sweden)

    O. V. Konoval

    2013-09-01

    Full Text Available The paper is devoted to the development and study of the mathematical model for corium melt interaction with low-temperature melting blocks in the passive protection systems (PPS against severe accidents at the NPP, and learning the peculiarities of construction and operation of the PPS. The configurations of cooling blocks’ distributions considered and the results of their work in the corium cooling pool are compared to the data of oth-er PPS’s conceptions. The conclusion is made that the models developed and the results obtained may be useful for constructing the PPS against severe accidents.

  11. Comparative study on aerosol removal by natural processes in containment in severe accident for AP1000 reactor

    International Nuclear Information System (INIS)

    Sun, Xiaohui; Cao, Xinrong; Shi, Xingwei; Yan, Jin

    2017-01-01

    Highlights: • Characteristics of aerosol distribution in containment are obtained. • Aerosol removal by natural processes is comparative studied by two methods. • Traditional rapid assessment method is conservative and can be applied in AP1000 reactor. - Abstract: Focusing on aerosol removal by naturally occurring processes in containment in severe accident for AP1000, integral severe accident code MELCOR and rapid assessment method mentioned in NUREG/CR-6189 are utilized to study aerosol removal by natural processes, respectively. Three typical severe accidents, induced by large break loss of coolant accident (LBLOCA), small break loss of coolant accident (SBLOCA) and steam generator tube rupture (SGTR), respectively, are selected for the study. The results obtained by two methods were further compared in the following several aspects: efficiency of aerosol removal by natural processes, peak time of aerosol suspended in containment atmosphere, peak amount of aerosol suspended in containment atmosphere, time when aerosol removal efficiency by natural processes is up to 99.9%. It was further concluded that results obtained by rapid assessment with shorter calculation process are more conservative. The analysis results provide reference to assessment method selection of severe accident source term for AP1000 nuclear emergency.

  12. The effect of omega-3 carboxylic acids on apolipoprotein CIII-containing lipoproteins in severe hypertriglyceridemia.

    Science.gov (United States)

    Morton, Allyson M; Furtado, Jeremy D; Lee, Jane; Amerine, William; Davidson, Michael H; Sacks, Frank M

    Lipoprotein subspecies containing apoCIII adversely affect cardiovascular disease (CVD) risk; for example, low density lipoprotein (LDL) with apoCIII is a stronger CVD predictor than LDL without apoCIII. The Epanova for Lowering Very High Triglycerides (EVOLVE) trial showed that Epanova (omega-3 carboxylic acids [OM3-CA]) significantly lowered TG and apoCIII but raised LDL-C. However, it is unknown what subspecies of LDL were affected by treatment. To determine how lipoprotein subspecies are affected by omega-3 fatty acid treatment, we studied the effect of OM3-CA on apoCIII concentrations in high density lipoprotein (HDL), LDL, and very low density lipoprotein (VLDL) and on the concentrations of subspecies of HDL, LDL, and VLDL that contain or do not contain apoCIII. We analyzed plasma from a subset of subjects from the EVOLVE trial, a 12-week double-blind study of 399 subjects with fasting TG of 500 to 2000 mg/dL who were randomized to OM3-CA 2, 3, or 4 g/d or olive oil (placebo). OM3-CA significantly reduced plasma apoCIII relative to placebo, as well as apoCIII in HDL, and apoCIII in LDL. Treatment did not significantly affect the concentration of LDL with apoCIII, a subspecies highly associated with CVD risk. OM3-CA increased selectively the concentration of LDL that does not contain apoCIII, a subspecies with a weak relation to coronary heart disease. The reduction in apoCIII was associated with plasma increases in eicosapentaenoic acid, docosahexaenoic acid, and arachidonic acid and decreases in linoleic, palmitic, and oleic acids. Reduction in apoCIII may be a mechanism for the TG-lowering effects of OM3-CA. The increase in LDL-C seen in the EVOLVE trial may not be associated with increased risk of CVD. Copyright © 2016 National Lipid Association. Published by Elsevier Inc. All rights reserved.

  13. Mobile/Modular BSL-4 Facilities for Meeting Restricted Earth Return Containment Requirements

    Science.gov (United States)

    Calaway, M. J.; McCubbin, F. M.; Allton, J. H.; Zeigler, R. A.; Pace, L. F.

    2017-01-01

    NASA robotic sample return missions designated Category V Restricted Earth Return by the NASA Planetary Protection Office require sample containment and biohazard testing in a receiving laboratory as directed by NASA Procedural Requirement (NPR) 8020.12D - ensuring the preservation and protection of Earth and the sample. Currently, NPR 8020.12D classifies Restricted Earth Return for robotic sample return missions from Mars, Europa, and Enceladus with the caveat that future proposed mission locations could be added or restrictions lifted on a case by case basis as scientific knowledge and understanding of biohazards progresses. Since the 1960s, sample containment from an unknown extraterrestrial biohazard have been related to the highest containment standards and protocols known to modern science. Today, Biosafety Level (BSL) 4 standards and protocols are used to study the most dangerous high-risk diseases and unknown biological agents on Earth. Over 30 BSL-4 facilities have been constructed worldwide with 12 residing in the United States; of theses, 8 are operational. In the last two decades, these brick and mortar facilities have cost in the hundreds of millions of dollars dependent on the facility requirements and size. Previous mission concept studies for constructing a NASA sample receiving facility with an integrated BSL-4 quarantine and biohazard testing facility have also been estimated in the hundreds of millions of dollars. As an alternative option, we have recently conducted an initial trade study for constructing a mobile and/or modular sample containment laboratory that would meet all BSL-4 and planetary protection standards and protocols at a faction of the cost. Mobile and modular BSL-2 and 3 facilities have been successfully constructed and deployed world-wide for government testing of pathogens and pharmaceutical production. Our study showed that a modular BSL-4 construction could result in approximately 90% cost reduction when compared to

  14. Development of the environmental qualification safety requirement matrix for the containment system of in-service CANDU reactors

    International Nuclear Information System (INIS)

    Chun, R.M.; Low, J.; Sobolewski, J.

    1994-01-01

    Over the last several years, Ontario Hydro Nuclear (OHN) has placed increasing emphasis on environmental qualification (EQ) at its Pickering and Bruce NGS A and B nuclear generating stations (NGSs). The program currently underway (at the time of the conference) builds upon the experience gained from the extensive Darlington NGS EQ experience and from EQ programs conducted by other utilities. Some of the major steps of the OHN EQ program include: defining Safety Requirement Matrices (SRMs), establishing environmental conditions, developing an EQ List, conducting an EQ Assessment and maintaining Operational EQ Assurance during the plant life. The SRM identifies safety related components, their required safety functions and their mission times for each postulated design basis accident (DBA). This is a critical step, as the SRM defines the equipment that requires assurance of EQ and precise requirements must be provided to ensure a cost effective EQ program. This paper describes the development of the SRMs for the containment system of the Bruce stations. The introductory section briefly discusses how the industry has dealt with equipment qualification as it has evolved and the role of the SRMs in the OHN EQ Program. In Section 2, the preparation of the SRM is described along with the applicable ground rules used. The results of the application of the SRM preparation guidelines to the containment system are discussed in Section 3. A summary of the major findings and conclusions is presented. 3 refs., 3 figs

  15. THAI experimental programme for containment safety assessment under severe accident conditions

    Energy Technology Data Exchange (ETDEWEB)

    Gupta, S.; Freitag, M. [Becker Technologies GmbH, Eschborn (Germany); Poss, G.

    2016-05-15

    The THAI (THAI = Thermal hydraulics, Hydrogen, Aerosols, Iodine) experimental programme aims to address open questions concerning the behavior of hydrogen, iodine and aerosols in the containment of water cooled reactors. Since its construction in 2000, THAI programme is being performed in the frame of various national projects (sponsored by German Federal Ministry for Economic Affairs and Energy, BMWi) and two international joint projects (under auspices of OECD/NEA). THAI experimental data have been widely used for the validation and further development of Lumped Parameter (LP) and Computational Fluid Dynamics (CFD) codes with 3D capabilities. Selected examples of code benchmark exercises performed based on the THAI data include; hydrogen distribution experiment (ISP-47 and OECD/NEA THAI code benchmark), hydrogen combustion behaviour (ISP-49), hydrogen mitigation by PARs (OECD/NEA THAI-2 code benchmark), iodine/surface interactions, iodine mass transfer, and iodine transport and multi-compartment behaviour (EU-SARNET and EU-SARNET2), thermal-hydraulic tests (German CFD-network). In the present paper, a brief overview on the THAI experiments and their role in the containment safety assessment is discussed.

  16. Quality assurance procedures for the CONTAIN severe reactor accident computer code

    International Nuclear Information System (INIS)

    Russell, N.A.; Washington, K.E.; Bergeron, K.D.; Murata, K.K.; Carroll, D.E.; Harris, C.L.

    1991-01-01

    The CONTAIN quality assurance program follows a strict set of procedures designed to ensure the integrity of the code, to avoid errors in the code, and to prolong the life of the code. The code itself is maintained under a code-configuration control system that provides a historical record of changes. All changes are incorporated using an update processor that allows separate identification of improvements made to each successive code version. Code modifications and improvements are formally reviewed and checked. An exhaustive, multilevel test program validates the theory and implementation of all codes changes through assessment calculations that compare the code-predicted results to standard handbooks of idealized test cases. A document trail and archive establish the problems solved by the software, the verification and validation of the software, software changes and subsequent reverification and revalidation, and the tracking of software problems and actions taken to resolve those problems. This document describes in detail the CONTAIN quality assurance procedures. 4 refs., 21 figs., 4 tabs

  17. Shipping container response to severe highway and railway accident conditions: Main report

    International Nuclear Information System (INIS)

    Fischer, L.E.; Chou, C.K.; Gerhard, M.A.; Kimura, C.Y.; Martin, R.W.; Mensing, R.W.; Mount, M.E.; Witte, M.C.

    1987-02-01

    This report describes a study performed by the Lawrence Livermore National Laboratory to evaluate the level of safety provided under severe accident conditions during the shipment of spent fuel from nuclear power reactors. The evaluation is performed using data from real accident histories and using representative truck and rail cask models that likely meet 10 CFR 71 regulations. The responses of the representative casks are calculated for structural and thermal loads generated by severe highway and railway accident conditions. The cask responses are compared with those responses calculated for the 10 CFR 71 hypothetical accident conditions. By comparing the responses it is determined that most highway and railway accident conditions fall within the 10 CFR 71 hypothetical accident conditions. For those accidents that have higher responses, the probabilities anf potential radiation exposures of the accidents are compared with those identified by the assessments made in the ''Final Environmental Statement on the Transportation of Radioactive Material by Air and other Modes,'' NUREG-0170. Based on this comparison, it is concluded that the radiological risks from spent fuel under severe highway and railway accident conditions as derived in this study are less than risks previously estimated in the NUREG-0170 document

  18. Experimental results from containment piping bellows subjected to severe accident conditions: Results from bellows tested in corroded conditions. Volume 2

    International Nuclear Information System (INIS)

    Lambert, L.D.; Parks, M.B.

    1995-10-01

    Bellows are an integral part of the containment pressure boundary in nuclear power plants. They are used at piping penetrations to allow relative movement between piping and the containment wall, while minimizing the load imposed on the piping and wall. Piping bellows are primarily used in steel containments; however, they have received limited use in some concrete (reinforced and prestressed) containments. In a severe accident they may be subjected to pressure and temperature conditions that exceed the design values, along with a combination of axial and lateral deflections. A test program to determine the leak-tight capacity of containment penetration bellows is being conducted at Sandia National Laboratories under the sponsorship of the US Nuclear Regulatory Commission. Several different bellows geometries, representative of actual containment bellows, have been subjected to extreme deflections along with pressure and temperature loads. The bellows geometries and loading conditions are described along with the testing apparatus and procedures. A total of nineteen bellows have been tested. Thirteen bellows were tested in ''like-new'' condition (results reported in Volume 1), and six were tested in a corroded condition. The tests showed that bellows in ''like-new'' condition are capable of withstanding relatively large deformations, up to, or near, the point of full compression or elongation, before developing leakage, while those in a corroded condition did not perform as well, depending on the amount of corrosion. The corroded bellows test program and results are presented in this report

  19. Prediction of hydrogen concentration in nuclear power plant containment under severe accidents using cascaded fuzzy neural networks

    Energy Technology Data Exchange (ETDEWEB)

    Choi, Geon Pil; Kim, Dong Yeong; Yoo, Kwae Hwan; Na, Man Gyun, E-mail: magyna@chosun.ac.kr

    2016-04-15

    Highlights: • We present a hydrogen-concentration prediction method in an NPP containment. • The cascaded fuzzy neural network (CFNN) is used in this prediction model. • The CFNN model is much better than the existing FNN model. • This prediction can help prevent severe accidents in NPP due to hydrogen explosion. - Abstract: Recently, severe accidents in nuclear power plants (NPPs) have attracted worldwide interest since the Fukushima accident. If the hydrogen concentration in an NPP containment is increased above 4% in atmospheric pressure, hydrogen combustion will likely occur. Therefore, the hydrogen concentration must be kept below 4%. This study presents the prediction of hydrogen concentration using cascaded fuzzy neural network (CFNN). The CFNN model repeatedly applies FNN modules that are serially connected. The CFNN model was developed using data on severe accidents in NPPs. The data were obtained by numerically simulating the accident scenarios using the MAAP4 code for optimized power reactor 1000 (OPR1000) because real severe accident data cannot be obtained from actual NPP accidents. The root-mean-square error level predicted by the CFNN model is below approximately 5%. It was confirmed that the CFNN model could accurately predict the hydrogen concentration in the containment. If NPP operators can predict the hydrogen concentration in the containment using the CFNN model, this prediction can assist them in preventing a hydrogen explosion.

  20. Early results from an experimental program to determine the behavior of containment piping penetration bellows subjected to severe accident conditions

    International Nuclear Information System (INIS)

    Lambert, L.D.; Parks, M.B.

    1994-01-01

    Containment piping penetration bellows are an integral part of the pressure boundary in steel containments in the United States (US). Their purpose is to minimize loading on the containment shell caused by differential movement between the piping and the containment. This differential movement is typically caused by thermal gradients generated during startup and shutdown of the reactor, but can be caused by earthquake, a loss-of-coolant accident (LOCA), or ''severe'' accidents. In the event of a severe accident, the bellows would be subjected to pressure, temperature, and deflection well beyond the design basis. Most bellows are installed such that they would be subjected to elevated internal pressure, elevated temperature, axial compression, and lateral deflection during a severe accident. A few bellows would be subjected to external pressure and axial elongation, as well as elevated temperature and lateral deflection. The purpose of this experimental program is to examine the potential for leakage of containment bellows during a severe accident. The test series subjects bellows to various levels and combinations of internal pressure, elevated temperature, axial compression or elongation, and lateral deformation. The experiments are being conducted in two parts. For Part 1, all bellows specimens are tested in ''like-new'' condition, without regard for the possible degrading effect of corrosion that has been observed in some containment piping bellows in the US Part I testing, which included 13 bellows tests, has been completed. The second part of the experimental program, in which bellows are subjected to simulated corrosive environments prior to testing, has just just begun. The Part I experiments have shown that bellows in ''like-new'' condition can withstand elevated temperatures and pressures along with large deformations before leaking. In most cases, the like-new bellows were fully compressed without developing any leakage

  1. An examination of source material requirements contained in 10 CFR Part 40

    International Nuclear Information System (INIS)

    Nussbaumer, D.; Smith, D.A.; Wiblin, C.

    1992-10-01

    This report identifies issues for consideration for rule-making to update the requirements for source material in 10 CFR Part 40 and examines options for resolving these issues. The contemplated rulemaking is intended to update 10 CFR Part 40 to reflect current radiation protection principles and regulatory practices. It is expected that such an update would make requirements for the control of source material more comparable to those pertaining to byproduct material contained in 10 CFR Part 30. The newer biological data and dose calculation methodology reflected in revised 10 CFR Part 20 will be used in analyses of potential regulatory amendments. This report presents historical background information and discussion on the various issues identified and makes preliminary recommendations concerning needed regulatory changes and approaches to rulemaking

  2. Predictability of iodine chemistry in the containment of a nuclear power plant under hypothetical severe accident conditions

    Energy Technology Data Exchange (ETDEWEB)

    Herranz, L.E.; Vela-Garcia, M.; Fontanet, J. [Unit of Nuclear Safety Research, CIEMAT, Madrid (Spain)

    2007-07-01

    One of the areas of top interest in the arena of severe accidents to get an accurate prediction of Source Term is Iodine Chemistry. In this paper an assessment of the current capability of MELCOR and ASTEC to predict iodine chemistry within containment in case of a postulated severe accident has been carried out. The experiments FPT1 and FPT2 of the PHEBUS-FP project have been used for comparisons, since they were carried out under rather different containment conditions during the chemistry phase (subcooled vs. saturated sump or acid vs. alkaline pH), which makes them very suitable to assess the current modeling capability of in-containment iodine chemistry models. The results obtained indicate that, even though, both integral codes have specific areas related to iodine chemistry that should be further developed and that their approach to the matter is drastically different, at present ASTEC-IODE allows for a more comprehensive simulation of the containment iodine chemistry. More importantly, lack of maturity of these codes would potentially maximize the so-called user-effect, so that it would be highly recommendable to perform sensitivity studies around iodine chemistry aspects when calculating Source Term scenarios. Key aspects needed of further research are: gaseous iodine chemistry (absent in MELCOR), organic iodine chemistry and adsorption/desorption on/from containment surfaces. (authors)

  3. Containment event analysis for postulated severe accidents: Peach Bottom Atomic Power Station, Unit 2. Draft report for comment

    Energy Technology Data Exchange (ETDEWEB)

    Amos, C N [Technadyne Engineering Consultants, Inc., Albuquerque, NM (United States); Griesmeyer, J M [Sandia National Laboratories, Albuquerque, NM (United States); Kolaczkowski, A M [Science Applications International Corporation, Albuquerque, NM (United States)

    1987-05-01

    A study has been performed as part of the Severe Accident Risk Reduction Program (SARRP) to investigate the response of a particular boiling water reactor with a Mark I containment (Peach Bottom Unit 2) to postulated severe accidents. A detailed containment event tree for the Peach Bottom plant has been developed to describe the various possible accident pathways that can lead to radioactive releases from containment. Data and analyses from a large number of NRC and industry-sponsored programs have been reviewed and used as a basis for quantifying the event tree, i.e., determining the likelihood of the pathways at each branch point for a variety of accident sequence initiators. A generalized containment event tree code, called EVNTRE, has been developed to facilitate the quantification. The uncertainty in the results has been examined by performing the quantification three times, using a different set of input each time to represent the variation of opinion in the reactor safety community. In the so-called 'central' estimate, the likelihood of early containment failure (occurring before or within a short time after reactor vessel breach) was found to be significant because of the possible occurrence of the following phenomena that can threaten containment integrity: (1) meltthrough of the drywell shell caused by thermal attack from core debris, and (2) drywell overpressurization caused by rapid depressurization of the reactor vessel in combination with other events such as direct heating. However, uncertainties surrounding these issues could cause the early failure likelihood to be significantly lower than in the central estimate. This work supports NRC's assessment of severe accident risks to be published in NUREG-1150. (author)

  4. Kinetics of Several Oxygen-Containing Carbon-Centered Free Radical Reactions with Nitric Oxide.

    Science.gov (United States)

    Rissanen, Matti P; Ihlenborg, Marvin; Pekkanen, Timo T; Timonen, Raimo S

    2015-07-16

    Kinetics of four carbon-centered, oxygen-containing free radical reactions with nitric oxide (NO) were investigated as a function of temperature at a few Torr pressure of helium, employing flow tube reactors coupled to a laser-photolysis/resonance-gas-discharge-lamp photoionization mass spectrometer (LP-RPIMS). Rate coefficients were directly determined from radical (R) decay signals under pseudo-first-order conditions ([R]0 ≪ [NO]). The obtained rate coefficients showed negative temperature dependences, typical for a radical-radical association process, and can be represented by the following parametrizations (all in units of cm(3) molecule(-1) s(-1)): k(CH2OH + NO) = (4.76 × 10(-21)) × (T/300 K)(15.92) × exp[50700/(RT)] (T = 266-363 K, p = 0.79-3.44 Torr); k(CH3CHOH + NO) = (1.27 × 10(-16)) × (T/300 K)(6.81) × exp[28700/(RT)] (T = 241-363 K, p = 0.52-3.43 Torr); k(CH3OCH2 + NO) = (3.58 ± 0.12) × 10(-12) × (T/300 K)(-3.17±0.14) (T = 221-363 K, p = 0.50-0.80 Torr); k(T)3 = 9.62 × 10(-11) × (T/300 K)(-5.99) × exp[-7100/(RT)] (T = 221-473 K, p = 1.41-2.95 Torr), with the uncertainties given as standard errors of the fits and the overall uncertainties estimated as ±20%. The rate of CH3OCH2 + NO reaction was measured in two density ranges due to its observed considerable pressure dependence, which was not found in the studied hydroxyalkyl reactions. In addition, the CH3CO + NO rate coefficient was determined at two temperatures resulting in k298K(CH3CO + NO) = (5.6 ± 2.8) × 10(-13) cm(3) molecule(-1) s(-1). No products were found during these experiments, reasons for which are briefly discussed.

  5. Monitoring Conformance and Containment for Geological Carbon Storage: Can Technology Meet Policy and Public Requirements?

    Science.gov (United States)

    Lawton, D. C.; Osadetz, K.

    2014-12-01

    The Province of Alberta, Canada identified carbon capture and storage (CCS) as a key element of its 2008 Climate Change strategy. The target is a reduction in CO2 emissions of 139 Mt/year by 2050. To encourage uptake of CCS by industry, the province has provided partial funding to two demonstration scale projects, namely the Quest Project by Shell and partners (CCS), and the Alberta Carbon Trunk Line Project (pipeline and CO2-EOR). Important to commercial scale implementation of CCS will be the requirement to prove conformance and containment of the CO2 plume injected during the lifetime of the CCS project. This will be a challenge for monitoring programs. The Containment and Monitoring Institute (CaMI) is developing a Field Research Station (FRS) to calibrate various monitoring technologies for CO2 detection thresholds at relatively shallow depths. The objective being assessed with the FRS is sensitivity for early detection of loss of containment from a deeper CO2 storage project. In this project, two injection wells will be drilled to sandstone reservoir targets at depths of 300 m and 700 m. Up to four observation wells will be drilled with monitoring instruments installed. Time-lapse surface and borehole monitoring surveys will be undertaken to evaluate the movement and fate of the CO2 plume. These will include seismic, microseismic, cross well, electrical resistivity, electromagnetic, gravity, geodetic and geomechanical surveys. Initial baseline seismic data from the FRS will presented.

  6. Isotope distributions in primary heat transport and containment systems during a severe accident in CANDU type reactor

    International Nuclear Information System (INIS)

    Constantin, M.

    2005-01-01

    The paper is intended to analyse the distribution of the fission products (FPs) in CANDU Primary Heat Transport (PHT) and CANDU Containment Systems by using the ASTEC code. The complexity of the data required by ASTEC and the complexity both of CANDU PHT and Containment System were strong motivation to begin with a simplified model. The data related to the nodes' definitions, temperatures and pressure conditions were chosen as possible as real data from CANDU loss of coolant accident sequence (CATHENA code results). The source term of FPs introduced into the PHT was estimated by ORIGEN code. The FPs distribution in the nodes of the circuit and the FPs mass transfer per isotope and chemical species were obtained by using SOPHAEROS module. The distributions within the containment are obtained by the CPA module (thermalhydraulic calculations in the containment and FPs aerosol transport). The results consist of mass distributions in the nodes of the circuit and the transferred mass to the containment through the break for different species (FPs and chemical species) and mass distributions in the different parts of the containment and different hosts. (authors)

  7. Prognostic significance of blood-brain barrier disruption in patients with severe nonpenetrating traumatic brain injury requiring decompressive craniectomy.

    Science.gov (United States)

    Ho, Kwok M; Honeybul, Stephen; Yip, Cheng B; Silbert, Benjamin I

    2014-09-01

    The authors assessed the risk factors and outcomes associated with blood-brain barrier (BBB) disruption in patients with severe, nonpenetrating, traumatic brain injury (TBI) requiring decompressive craniectomy. At 2 major neurotrauma centers in Western Australia, a retrospective cohort study was conducted among 97 adult neurotrauma patients who required an external ventricular drain (EVD) and decompressive craniectomy during 2004-2012. Glasgow Outcome Scale scores were used to assess neurological outcomes. Logistic regression was used to identify factors associated with BBB disruption, defined by a ratio of total CSF protein concentrations to total plasma protein concentration > 0.007 in the earliest CSF specimen collected after TBI. Of the 252 patients who required decompressive craniectomy, 97 (39%) required an EVD to control intracranial pressure, and biochemical evidence of BBB disruption was observed in 43 (44%). Presence of disruption was associated with more severe TBI (median predicted risk for unfavorable outcome 75% vs 63%, respectively; p = 0.001) and with worse outcomes at 6, 12, and 18 months than was absence of BBB disruption (72% vs 37% unfavorable outcomes, respectively; p = 0.015). The only risk factor significantly associated with increased risk for BBB disruption was presence of nonevacuated intracerebral hematoma (> 1 cm diameter) (OR 3.03, 95% CI 1.23-7.50; p = 0.016). Although BBB disruption was associated with more severe TBI and worse long-term outcomes, when combined with the prognostic information contained in the Corticosteroid Randomization after Significant Head Injury (CRASH) prognostic model, it did not seem to add significant prognostic value (area under the receiver operating characteristic curve 0.855 vs 0.864, respectively; p = 0.453). Biochemical evidence of BBB disruption after severe nonpenetrating TBI was common, especially among patients with large intracerebral hematomas. Disruption of the BBB was associated with more severe

  8. Disposal of TRU Waste from the PFP in pipe overpack containers to WIPP Including New Security Requirements

    International Nuclear Information System (INIS)

    HOPKINS, A.M.

    2003-01-01

    The Department of Energy is responsible for the safe management and cleanup of the DOE complex. As part of the cleanup and closure of the Plutonium Finishing Plant (PFP) located on the Hanford site, the nuclear material inventory was reviewed to determine the appropriate disposition path. Based on the nuclear material characteristics, the material was designated for stabilization and packaging for long term storage and transfer to the Savannah River Site, or a decision for discard was made. The discarded material was designated as waste material and slated for disposal to the Waste Isolation Pilot Plant (WIPP). Prior to preparing any residue wastes for disposal at the WIPP, several major activities need to be completed. As detailed a processing history as possible of the material including origin of the waste must be researched and documented. A technical basis for termination of safeguards on the material must be prepared and approved. Utilizing process knowledge and processing history, the material must be characterized, sampling requirements determined, acceptable knowledge package and waste designation completed prior to disposal. All of these activities involve several organizations including the contractor, DOE, state representatives and other regulators such as EPA. At PFP, a process has been developed for meeting the many, varied requirements and successfully used to prepare several residue waste streams including Rocky Flats incinerator ash, hanford incinerator ash and Sand, Slag and Crucible (SS and C) material for disposal. These waste residues are packed into Pipe Overpack Containers for shipment to the WIPP

  9. Full blood count and haemozoin-containing leukocytes in children with malaria: diagnostic value and association with disease severity

    Directory of Open Access Journals (Sweden)

    Lell Bertrand

    2008-06-01

    Full Text Available Abstract Background Diligent and correct laboratory diagnosis and up-front identification of risk factors for progression to severe disease are the basis for optimal management of malaria. Methods Febrile children presenting to the Medical Research Unit at the Albert Schweitzer Hospital (HAS in Lambaréné, Gabon, were assessed for malaria. Giemsa-stained thick films for qualitative and quantitative diagnosis and enumeration of malaria pigment, or haemozoin (Hz-containing leukocytes (PCL were performed, and full blood counts (FBC were generated with a Cell Dyn 3000® instrument. Results Compared to standard light microscopy of Giemsa-stained thick films, diagnosis by platelet count only, by malaria pigment-containing monocytes (PCM only, or by pigment-containing granulocytes (PCN only yielded sensitivities/specificities of 92%/93%; 96%/96%; and 85%/96%, respectively. The platelet count was significantly lower in children with malaria compared to those without (p ® instrument detected significantly more patients with PCL (p Conclusion In the age group examined in the Lambaréné area, platelets are an excellent adjuvant tool to diagnose malaria. Pigment-containing leukocytes (PCL are more readily detected by automated scatter flow cytometry than by microscopy. Automated Hz detection by an instrument as used here is a reliable diagnostic tool and correlates with disease severity. However, clinical usefulness as a prognostic tool is limited due to an overlap of PCL numbers recorded in severe versus non-severe malaria. However, this is possibly because of the instrument detection algorithm was not geared towards this task, and data lost during processing; and thus adjusting the instrument's algorithm may allow to establish a meaningful cut-off value.

  10. 10 CFR 32.22 - Self-luminous products containing tritium, krypton-85 or promethium-147: Requirements for license...

    Science.gov (United States)

    2010-01-01

    ... 10 Energy 1 2010-01-01 2010-01-01 false Self-luminous products containing tritium, krypton-85 or... containing tritium, krypton-85 or promethium-147: Requirements for license to manufacture, process, produce... self-luminous products containing tritium, krypton-85, or promethium-147, or to initially transfer such...

  11. Performance of the primary containment of a BWR during a severe accident whit the code RELAP/SCDAPSIM

    International Nuclear Information System (INIS)

    Castillo G, F.

    2015-01-01

    In this thesis work, it was developed a model of the vacuum breaker valves and down comers for a BWR Mark II primary containment for the code RELAP/SCDAPSIM Mod. 3.4. This code was used to simulate a Station Blackout (Sbo) that evolves to a severe accident scenario. To accomplish this task, the vacuum breaker valves and down comers were included in a simplified model of the primary containment that includes both wet well and dry well, which was coupled with a model of the Nuclear Steam Supply System (NSSS), in order to study the behavior of the primary containment during the evolution of the accident scenario. In the analysis of the results of the simulation, the behavior of the wet well and dry well during the event was particularly monitored, by analyzing the evolution of temperature and pressure profiles in such volumes, this to determine the impact of the inclusion of the breaker vacuum valves and down comers. The results show that the effect of this extension of the model is that more conservative results are obtained, i.e., higher pressures are reached in both wet well and dry well than when it is used a containment model that does not include neither the vacuum valves nor the down comers. The most relevant results obtained show that the Rcic alone is able to keep the core fully covered, but even in such a case, it evaporates about 15% of the initial inventory of liquid water in the Pressure Suppression Pool (Psp). When the Rcic operation is lost, 20% more of the liquid water inventory in the Psp is further reduced within four to twelve hours (approximately), time at which the simulation crashed. Besides, there is a significant increase of pressure in the containment. As the accident evolves, the pressure in the containment continues increasing, but there is still considerable margin to reach the design pressure of the containment. At the end of the simulation, the results show a gauge pressure value of 224,550 Pa in the Psp and 187,482 Pa in the wet well

  12. A Giant Brunneroma Causing Gastrointestinal Bleeding and Severe Anemia Requiring Transfusion and Surgery

    Directory of Open Access Journals (Sweden)

    Nicola C. Frenkel

    2017-01-01

    Full Text Available Brunner’s gland hamartoma, also called hyperplasia, adenoma, and Brunneroma, is an extremely rare benign proliferative lesion of Brunner’s glands in the duodenum. While being mostly small and asymptomatic, they can result in gastrointestinal bleeding and obstruction. We report the case of a 54-year-old man presenting with melena and severe anemia requiring blood transfusion. CT scans showed a large mass of 8 cm in diameter, presumably arising in the duodenum. Endoscopic biopsies were not conclusive. As we were unable to determine the nature of the mass preoperatively and due to the severe symptoms, its size, and the uncertain malignant potential, a classic Whipple procedure was performed. The resected specimen showed extensive proliferation of Brunner’s glands without signs of malignancy.

  13. Acute respiratory failure requiring mechanical ventilation in severe chronic obstructive pulmonary disease (COPD).

    Science.gov (United States)

    Gadre, Shruti K; Duggal, Abhijit; Mireles-Cabodevila, Eduardo; Krishnan, Sudhir; Wang, Xiao-Feng; Zell, Katrina; Guzman, Jorge

    2018-04-01

    There are limited data on the epidemiology of acute respiratory failure necessitating mechanical ventilation in patients with severe chronic obstructive pulmonary disease (COPD). The prognosis of acute respiratory failure requiring invasive mechanical ventilation is believed to be grim in this population. The purpose of this study was to illustrate the epidemiologic characteristics and outcomes of patients with underlying severe COPD requiring mechanical ventilation.A retrospective study of patients admitted to a quaternary referral medical intensive care unit (ICU) between January 2008 and December 2012 with a diagnosis of severe COPD and requiring invasive mechanical ventilation for acute respiratory failure.We evaluated 670 patients with an established diagnosis of severe COPD requiring mechanical ventilation for acute respiratory failure of whom 47% were male with a mean age of 63.7 ± 12.4 years and Acute physiology and chronic health evaluation (APACHE) III score of 76.3 ± 27.2. Only seventy-nine (12%) were admitted with a COPD exacerbation, 27(4%) had acute respiratory distress syndrome (ARDS), 78 (12%) had pneumonia, 78 (12%) had sepsis, and 312 (47%) had other causes of respiratory failure, including pulmonary embolism, pneumothorax, etc. Eighteen percent of the patients received a trial of noninvasive positive pressure ventilation. The median duration of mechanical ventilation was 3 days (interquartile range IQR 2-7); the median duration for ICU length of stay (LOS) was 5 (IQR 2-9) days and the median duration of hospital LOS was 12 (IQR 7-22) days. The overall ICU mortality was 25%. Patients with COPD exacerbation had a shorter median duration of mechanical ventilation (2 vs 4 days; P = .04), ICU (3 vs 5 days; P = .01), and hospital stay (10 vs 13 days; P = .01). The ICU mortality (9% vs 27%; P respiratory failure. A 1-unit increase in the APACHE III score was associated with a 1% decrease and having an active cancer was associated

  14. Structural requirements for cub domain containing protein 1 (CDCP1 and Src dependent cell transformation.

    Directory of Open Access Journals (Sweden)

    Gwendlyn Kollmorgen

    Full Text Available Cub domain containing protein 1 (CDCP1 is strongly expressed in tumors derived from lung, colon, ovary, or kidney. It is a membrane protein that is phosphorylated and then bound by Src family kinases. Although expression and phosphorylation of CDCP1 have been investigated in many tumor cell lines, the CDCP1 features responsible for transformation have not been fully evaluated. This is in part due to the lack of an experimental system in which cellular transformation depends on expression of exogenous CDCP1 and Src. Here we use retrovirus mediated co-overexpression of c-Src and CDCP1 to induce focus formation of NIH3T3 cells. Employing different mutants of CDCP1 we show that for a full transformation capacity, the intact amino- and carboxy-termini of CDCP1 are essential. Mutation of any of the core intracellular tyrosine residues (Y734, Y743, or Y762 abolished transformation, and mutation of a palmitoylation motif (C689,690G strongly reduced it. Src kinase binding to CDCP1 was not required since Src with a defective SH2 domain generated even more CDCP1 dependent foci whereas Src myristoylation was necessary. Taken together, the focus formation assay allowed us to define structural requirements of CDCP1/Src dependent transformation and to characterize the interaction of CDCP1 and Src.

  15. A study on the hydrogen behavior and its mitigation in the APR1400 containment during a severe accident

    International Nuclear Information System (INIS)

    Kim, Jong Tae; Hong, Seong Wan; Park, Rae Joon; Kim, Sang Baik

    2005-02-01

    During a hypothetical severe accident in a nuclear power plant (NPP), hydrogen is generated by the active reaction of fuel-cladding and steam in the reactor pressure vessel and released with steam into the containment. In this study, the analysis of the hydrogen and steam behavior during selected severe accidents in the APR1400 containment has been conducted by using the GASFLOW code. For the SBLOCA, hydrogen was accumulated in the containment dome region quickly when only PARSs were used. When the igniters were turned on, a standing flame was formed around a coolant pump and burnt most of the hydrogen blown-out. For the TLOFW accident, the flap-type pressure damper installed at the IRWST vents strongly affected the flow structure of the hydrogen. And by the steam-rich and oxygen starvation conditions in the IRWST, DDT is not likely to occur. For the SBO accident, dry hydrogen was release in the IRWST by the assumption of full condensation of the released steam in the IRWST water. In this case, the possibility of flame acceleration is high in the IRWST and annular compartment. In this study two design modifications were proposed in view of the hydrogen mitigation strategy and their effectiveness was evaluated by the GASFLOW analysis

  16. Outcomes from the EURATOM–ROSATOM ERCOSAM SAMARA projects on containment thermal-hydraulics for severe accident management

    Energy Technology Data Exchange (ETDEWEB)

    Paladino, Domenico, E-mail: domenico.paladino@psi.ch [Paul Scherrer Institut (Switzerland); Andreani, Michele [Paul Scherrer Institut (Switzerland); Guentay, Salih [Innovative, Technology Development GmbH (Switzerland); Mignot, Guillaume; Kapulla, Ralf; Paranjape, Sidharth; Sharabi, Medhat [Paul Scherrer Institut (Switzerland); Kisselev, Arkadi; Yudina, Tatiana; Filippov, Aleksandr [Nuclear Safety Institute of the Russian Academy of Sciences, Moscow 115191 (Russian Federation); Kamnev, Mikhail; Khizbullin, Akhmir; Tyurikov, Oleg [JSC “Afrikantov OKB Mechanical Engineering”, Nizhny Novgorod 603074 (Russian Federation); Liang, Zhe [CNL-2251 Speakman Drive, Mississauga, ON L5K 1B2 (Canada); Abdo, Daniele; Brinster, Jérôme; Dabbene, Frédéric [CEA, DEN, DM2S, STMF, F-91191 Gif-sur-Yvette Cedex (France); Kelm, Stephan [Forschungszentrum Juelich, 52425 Jülich (Germany); Klauck, Michael; Götz, Lasse [RWTH Aachen University (Germany); and others

    2016-11-15

    Highlights: • Hydrogen distribution in the containment of PWR was investigated for scenario leading to stratification. • The scenario was scaled from a generic PWR containment to four facilities. • Effect of spray, cooler and heat sources was investigated experimentally and with LP and CFD. • Code-to-code benchmarks aiming a scaling up the facilities to a large containment. - Abstract: ERCOSAM and SAMARA are the acronyms for two parallel projects co-financed respectively by EURATOM and ROSATOM during the period 2010–2014 with the general aim to advance the knowledge on the phenomenology associated with the hydrogen and steam spreading and stratification in the LWR containment during a postulated severe accident. The important peculiarity of the projects was in experimental and analytical investigating the impact of systems such as spray, cooler and heat sources (simulating thermal effect of PARs) on the distribution of gas mixture (e.g. hydrogen, steam, air). This paper presents the main outcomes of the ERCOSAM–SAMARA projects.

  17. Outcomes from the EURATOM–ROSATOM ERCOSAM SAMARA projects on containment thermal-hydraulics for severe accident management

    International Nuclear Information System (INIS)

    Paladino, Domenico; Andreani, Michele; Guentay, Salih; Mignot, Guillaume; Kapulla, Ralf; Paranjape, Sidharth; Sharabi, Medhat; Kisselev, Arkadi; Yudina, Tatiana; Filippov, Aleksandr; Kamnev, Mikhail; Khizbullin, Akhmir; Tyurikov, Oleg; Liang, Zhe; Abdo, Daniele; Brinster, Jérôme; Dabbene, Frédéric; Kelm, Stephan; Klauck, Michael; Götz, Lasse

    2016-01-01

    Highlights: • Hydrogen distribution in the containment of PWR was investigated for scenario leading to stratification. • The scenario was scaled from a generic PWR containment to four facilities. • Effect of spray, cooler and heat sources was investigated experimentally and with LP and CFD. • Code-to-code benchmarks aiming a scaling up the facilities to a large containment. - Abstract: ERCOSAM and SAMARA are the acronyms for two parallel projects co-financed respectively by EURATOM and ROSATOM during the period 2010–2014 with the general aim to advance the knowledge on the phenomenology associated with the hydrogen and steam spreading and stratification in the LWR containment during a postulated severe accident. The important peculiarity of the projects was in experimental and analytical investigating the impact of systems such as spray, cooler and heat sources (simulating thermal effect of PARs) on the distribution of gas mixture (e.g. hydrogen, steam, air). This paper presents the main outcomes of the ERCOSAM–SAMARA projects.

  18. An optimal hydrogen control analysis for the in-containment refueling storage tank (IRWST) of the Korean next generation reactor (KNGR) containment under severe accidents

    International Nuclear Information System (INIS)

    Byung-Chul, Lee; Hee-Jin, Ko; Se-Won, Lee

    2001-01-01

    Under severe accidents that a large amount of hydrogen is expected to release, the In-Containment Refueling Water Storage Tank (IRWST) air space has more worse condition with respect to the hydrogen control since, as one of hydrogen source compartment, normally it is separated from the other compartments and has relatively small volume. The hydrogen concentrations in the IRWST gas space, when the hydrogen was directly released into this area, were analyzed using the MAAP4 code in order to investigate if locally very high concentrations could be reduced so that inadvertent detonation or detonation-to-deflagration (DDT) in this area might be prevented. For this purpose, the thermo-hydraulic and combustion phenomena being capable of occurring in the IRWST were also considered. As a result of numerical calculations with 12-compartment containment model, the time duration that the flammable gas mixture was formed was greatly decreased via oxygen-starved or steam-rich conditions, although instantaneously peak concentration itself could not be avoided. Moreover, if the diffusion flame or steam stripping can be occurred in the IRWST, it was expected to have more chance to control the hydrogen in the IRWST gas space. After the hydrogen finished to be rapidly released, the hydrogen in this area could be controlled by the PARs' hydrogen depletion and by igniter's deliberate burning. Especially, the review on the analyses for two typical, but most probable sequences of quite a different hydrogen release modes gives an insight that the flammable gas mixture in the IRWST can be avoid by rapid depressurization operation, which is recommendable for being implemented into accident management program. (authors)

  19. Regulatory requirements for the use of consumer products containing radioactive substances

    International Nuclear Information System (INIS)

    Mason, G.C.; Paynter, R.A.; Schmitt-Hannig, A.; Sztanyik, L.B.

    1996-01-01

    In almost 100 years since the discovery of radioactivity, the properties of radioactive materials have been exploited in products such as clocks and watches incorporating luminous paint which are freely available to members of the public. Over time, regulatory authorities have felt it necessary to apply some degree of control to the supply and use of such products in order to protect public health. In many areas of radiation protection, national authorities take note of international recommendations when developing national standards, but the existing detailed guidance of the International Atomic Energy Agency (IAEA) for consumer products is incomplete and out of date. Recently, a thorough revision of the International Basic Safety Standards (BSS) has occurred, which has prompted a review and revision of the related guidance published by the IAEA. A draft Guide on Regulatory Requirements for the Use of Consumer Products Containing Radioactive Substances has now been completed and is currently under review within the IAEA's system for development of documents in its Safety Series of publications. (author)

  20. Distribution of hydrogen within the HDR-containment under severe accident conditions. OECD standard problem. Final comparison report

    International Nuclear Information System (INIS)

    Karwat, H.

    1992-08-01

    The present report summarizes the results of the International Standard Problem Exercise ISP-29, based on the HDR Hydrogen Distribution Experiment E11.2. Post-test analyses are compared to experimentally measured parameters, well-known to the analysis. This report has been prepared by the Institute for Reactor Dynamics and Reactor Safety of the Technical University Munich under contract with the Gesellschaft fuer Anlagen- und Reaktorsicherheit (GRS) which received funding for this activity from the German Ministry for Research and Technology (BMFT) under the research contract RS 792. The HDR experiment E11.2 has been performed by the Kernforschungszentrum Karlsruhe (KfK) in the frame of the project 'Projekt HDR-Sicherheitsprogramm' sponsored by the BMFT. Ten institutions from eight countries participated in the post-test analysis exercise which was focussing on the long-lasting gas distribution processes expected inside a PWR containment under severe accident conditions. The gas release experiment was coupled to a long-lasting steam release into the containment typical for an unmitigated small break loss-of-coolant accident. In lieu of pure hydrogen a gas mixture consisting of 15% hydrogen and 85% helium has been applied in order to avoid reaching flammability during the experiment. Of central importance are common overlay plots comparing calculated transients with measurements of the global pressure, the local temperature-, steam- and gas concentration distributions throughout the entire HDR containment. The comparisons indicate relatively large margins between most calculations and the experiment. Having in mind that this exercise was specified as an 'open post-test' analysis of well-known measured data the reasons for discrepancies between measurements and simulations were extensively discussed during a final workshop. It was concluded that analytical shortcomings as well as some uncertainties of experimental boundary conditions may be responsible for deviations

  1. Distribution of hydrogen within the HDR-containment under severe accident conditions. OECD standard problem. Final comparison report

    Energy Technology Data Exchange (ETDEWEB)

    Karwat, H

    1992-08-15

    The present report summarizes the results of the International Standard Problem Exercise ISP-29, based on the HDR Hydrogen Distribution Experiment E11.2. Post-test analyses are compared to experimentally measured parameters, well-known to the analysis. This report has been prepared by the Institute for Reactor Dynamics and Reactor Safety of the Technical University Munich under contract with the Gesellschaft fuer Anlagen- und Reaktorsicherheit (GRS) which received funding for this activity from the German Ministry for Research and Technology (BMFT) under the research contract RS 792. The HDR experiment E11.2 has been performed by the Kernforschungszentrum Karlsruhe (KfK) in the frame of the project 'Projekt HDR-Sicherheitsprogramm' sponsored by the BMFT. Ten institutions from eight countries participated in the post-test analysis exercise which was focussing on the long-lasting gas distribution processes expected inside a PWR containment under severe accident conditions. The gas release experiment was coupled to a long-lasting steam release into the containment typical for an unmitigated small break loss-of-coolant accident. In lieu of pure hydrogen a gas mixture consisting of 15% hydrogen and 85% helium has been applied in order to avoid reaching flammability during the experiment. Of central importance are common overlay plots comparing calculated transients with measurements of the global pressure, the local temperature-, steam- and gas concentration distributions throughout the entire HDR containment. The comparisons indicate relatively large margins between most calculations and the experiment. Having in mind that this exercise was specified as an 'open post-test' analysis of well-known measured data the reasons for discrepancies between measurements and simulations were extensively discussed during a final workshop. It was concluded that analytical shortcomings as well as some uncertainties of experimental boundary conditions may be responsible for deviations

  2. 16 CFR 316.4 - Requirement to place warning labels on commercial electronic mail that contains sexually oriented...

    Science.gov (United States)

    2010-01-01

    ... commercial electronic mail that contains sexually oriented material. 316.4 Section 316.4 Commercial Practices FEDERAL TRADE COMMISSION REGULATIONS UNDER SPECIFIC ACTS OF CONGRESS CAN-SPAM RULE § 316.4 Requirement to place warning labels on commercial electronic mail that contains sexually oriented material. (a) Any...

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

    International Nuclear Information System (INIS)

    Baratta, A.J.

    1997-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

    Baratta, A.J.

    1997-07-01

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

  5. Automated container transportation using self-guided vehicles: Fernald site requirements

    International Nuclear Information System (INIS)

    Hazen, F.B.

    1993-09-01

    A new opportunity to improve the safety and efficiency of environmental restoration operations, using robotics has emerged from advances in industry, academia, and government labs. Self-Guided Vehicles (SGV's) have recently been developed in industry and early systems have already demonstrated much, though not all, of the functionality necessary to support driverless transportation of waste within and between processing facilities. Improved materials databases are being developed by at least two DOE remediation sites, the Fernald Environmental Management Project (FEME) in the State of Ohio and the Hanford Complex in the State of Washington. SGV's can be developed that take advantage of the information in these databases and yield improved dispatch, waste tracking, report and shipment documentation. In addition, they will reduce the radiation hazard to workers and the risk of damaging containers through accidental collision. In this document, features of remediation sites that dictate the design of both the individual SGV's and the collective system of SGV's are presented, through the example of the site requirements at Fernald. Some concepts borrowed from the world of manufacturing are explained and then used to develop an integrated, holistic view of the remediation site as a pseudo-factory. Transportation methods at Fernald and anticipated growth in transport demand are analyzed. The new site-wide database under development at Fernald is presented so that advantageous and synergistic links between SGV's and information systems can be analyzed. Details of the SGV development proposed are submitted, and some results of a recently completed state of the art survey for SGV use in this application are also presented

  6. Requirements for the coatings of a nuclear power plant containment building

    International Nuclear Information System (INIS)

    Orantie, K.; Kuosa, H.; Haekkae-Roennholm, E.

    2001-06-01

    The report presents the criteria for the inside coatings of nuclear power plant containment buildings including: radiation resistance, decontamination, chemical resistance in accident situations and fire resistance

  7. Severe HDN due to anti-Ce that required exchange tranfusion.

    Science.gov (United States)

    Wagner, T; Resch, B; Legler, T J; Mossier, C; Helmberg, W; Köhler, M; Lanzer, G

    2000-05-01

    Rh system antibodies are commonly encountered in blood bank practice as well as during pregnancy. Nevertheless, no examples of anti-Ce (RH7) have been reported as a cause of HDN that requires exchange transfusion. A 38-year-old woman in her fourth pregnancy was typed as blood group O D+, C-, c+, E+, e-. Anti-C and anti-e were detected in her serum during a routine prenatal work-up. Further evaluation, including flow cytometric analysis, revealed the presence of a strong anti-Ce and a weak anti-e. Her partner was typed as group A D+, C+, c-, E-, e+. A seemingly healthy male infant was delivered at 40 weeks of gestation. The infant's RBCs were typed as group O D-, C+, c+, E+, e+ with a positive DAT (titer 128). Twenty-five hours after birth, the baby had to be transferred to the neonatal intensive care unit because of rapidly rising total serum bilirubin. Despite intensive treatment, including double phototherapy, albumin infusion, and the administration of furosemide and IVIG, the total serum bilirubin level increased during the following day and exchange transfusion with 2 units of type O D-, C-, c+, E+, e- had to be performed; this resulted in a prompt decrease in total serum bilirubin without relapse. Anti-Ce caused severe HDN requiring exchange transfusion. This highlights the need for a close follow-up throughout pregnancy if unexpected RBC antibodies are present, to permit the provision of compatible blood in case of a rare antibody.

  8. Comparison of european computer codes relative to the aerosol behavior in PWR containment buildings during severe core damage accidents

    International Nuclear Information System (INIS)

    Fermandjian, J.; Beonio-Brocchieri, F.

    1986-09-01

    The present study concerns a comparative exercise, performed within the framework of the Commission of the European Communities, of the computer codes used in reactor safety in order to assess their capability of realistically describing the aerosol behavior in PWR reactor containment buildings during severe accidents. The codes included in the present study are the following: AEROSIM-M, AEROSOLS/Bl, CORRAL-2, NAUA Mod5. In AEROSIM-M, AEROSOLS/Bl and NAUA Mod5, the integro-differential equation for the evolution of the particle mass distribution is approximated by a set of coupled first order differential equations. To this end, the particle distribution function is replaced by a number of discrete monodisperse fractions. The CORRAL-2 has an essentially empirical basis (processes not explicitely modelled, but their net effects accounted for). The physical processes taken into account in the codes are shown finally

  9. Study of the ruthenium fission-product behavior in the containment, in the case of a nuclear reactor severe accident

    International Nuclear Information System (INIS)

    Mun, Ch.

    2007-03-01

    Ruthenium tetroxide is an extremely volatile and highly radio-toxic species. During a severe accident with air ingress in the reactor vessel, ruthenium oxides may reach the reactor containment building in significant quantities. Therefore, a better understanding of the RuO 4 (g) behaviour in the containment atmosphere is of primary importance for the assessment of radiological consequences, in the case of potential releases of this species into the environment. A RuO 4 (g) decomposition kinetic law was determined. Steam seems to play a catalytic role, as well as the presence of ruthenium dioxide deposits. The temperature is also a key parameter. The nature of the substrate, stainless steel or paint, did not exhibit any chemical affinities with RuO 4 (g). This absence of reactivity was confirmed by XPS analyses, which indicate the presence of the same species in the Ru deposits surface layer whatever the substrates considered. It has been concluded that RuO 4 (g) decomposition corresponds to a bulk gas phase decomposition. The ruthenium re-volatilization phenomenon under irradiation from Ru deposits was also highlighted. An oxidation kinetic law was determined. The increase of the temperature and the steam concentration promote significantly the oxidation reaction. The establishment of Ru behavioural laws allowed making a modelling of the Ru source term. The results of the reactor calculations indicate that the values obtained for 106 Ru source term are closed to the reference value considered currently by the IRSN, for 900 MWe PWR safety analysis. (author)

  10. 40 CFR 267.195 - What are the secondary containment requirements?

    Science.gov (United States)

    2010-07-01

    ... wastes or accumulated liquid out of the system to the soil, groundwater, or surface water at any time... the failure of either the primary or secondary containment structure or the presence of any release of... 40 Protection of Environment 26 2010-07-01 2010-07-01 false What are the secondary containment...

  11. Salovum egg yolk containing antisecretory factor as an adjunct therapy in severe cholera in adult males: a pilot study.

    Science.gov (United States)

    Alam, Nur H; Ashraf, Hasan; Olesen, Maryam; Salam, Mohammed A; Gyr, Niklaus; Meier, Remy

    2011-08-01

    Cholera involves stimulation of intestinal secretory process in response to cholera toxin leading to profuse watery diarrhoea that might cause death due to dehydration unless timely rehydration therapy is initiated. Efforts to identify and test potential antisecretory agents are ongoing. Antisecretory factor (AF) is a naturally-occurring protein produced in the human secretory organs, including the intestine, with antisectory properties demonstrated in animal and human models of secretory diarrhoea. Salovum egg yolk powder contains antisecretory proteins in a much higher (500 times) concentration than that of normal hen eggs. This is achieved by feeding hens with specially-processed cereals, capable of inducing antisecretory proteins in the yolk. The aim of the study was to examine the effect of Salovum egg yolk powder containing AF in the treatment of adult cholera patients. In an open, randomized controlled trial (pilot study), 40 adult male patients with severe cholera were studied: 20 received standard treatment (oral rehydration solution, antibiotic, and usual hospital diet) plus Salovum egg yolk powder (study group) and 20 received standard treatment alone (control group). All the patients received tablet doxycycline (300 mg) once immediately after randomization. Written informed consent was obtained from each subject before enrollment. The main outcome measures were stool weight and duration of diarrhoea. The demographic and baseline clinical characteristics of the study patients were comparable between the groups. No significant differences were found in the mean stool weight, g/kg of body-weight during the first 24 hours [study vs control group, mean +/- standard deviation (SD), 218 +/- 119 vs 195 +/- 136], second 24 hours (mean +/- SD, 23 +/- 39 vs 22 +/- 34), and cumulative up to 72 hours (mean +/- SD, 245 +/- 152 vs 218 +/- 169). The duration (hours) of diarrhoea after admission in the hospital was also similar in both the groups (mean +/- SD, 33 +/- 14

  12. Study on the design and manufacturing requirements of container for low level radioactive solid waste form KRR decommissioning

    International Nuclear Information System (INIS)

    Lee, D. K.; Kim, H. R.; Park, S. K.; Jung, K. H.; Jung, W. S.; Jung, K. J.

    2000-01-01

    The design requirement and manufacturing criteria have been proposed on the container for the storage and transportation of low level radioactive solid waste from decommissioning of KRR 1 and 2. The structure analysis was carried out based on the design criteria, and the safety of the container was assessed. The ISO container with its capacity of 4m 3 was selected for the radioactive solid waste storage. The proposed container was satisfied the criteria of ISO 1496/1 and the packaging standard of atomic energy act. manufacturing and test standards of IAEA were also applied to the container. Stress distribution and deformation were analyzed under given condition using ANSYS code, and the maximum stress was verified to be within yield stress without any structural deformation. From the results of lifting tests, it was verified that the container was safe

  13. Mild Caloric Restriction Decreases Insulin Requirements in Patients With Type 2 Diabetes and Severe Insulin Resistance.

    Science.gov (United States)

    Meehan, Cristina Adelia; Cochran, Elaine; Mattingly, Megan; Gorden, Phillip; Brown, Rebecca J

    2015-07-01

    Type 2 diabetes (T2D) affects ~10% of the US population, a subset of whom have severe insulin resistance (SIR) (>200 units/d). Treatment of these patients with high-dose insulin presents logistical and compliance challenges. We hypothesized that mild caloric restriction would reduce insulin requirements in patients with T2D and SIR.This was a retrospective study at the National Institutes of Health Clinical Center. Inclusion criteria were as follows: T2D, and insulin dose >200 units/d or >2 units/kg/d. The intervention consisted of mild caloric restriction during a 3 to 6-day hospitalization. The major outcomes were change in insulin dose and blood glucose from admission to discharge.Ten patients met inclusion criteria. Baseline glycated hemoglobin A1c was 10.0 ± 1.6% and body mass index 38.8 ± 9.0 kg/m. Food intake was restricted from 2210 ± 371 kcal/d preadmission to 1810 ± 202 during the hospital stay (16.5% reduction). Insulin dose decreased from 486 ± 291 units/d preadmission to 223 ± 127 at discharge (44% reduction, P = 0.0025). Blood sugars decreased nonsignificantly in the fasting state (from 184 ± 85 to 141 ± 42, P = 0.20), before lunch (239 ± 68 to 180 ± 76, P = 0.057), and at bedtime (212 ± 95 to 176 ± 48, P = 0.19), and significantly decreased before dinner (222 ± 92 to 162 ± 70, P = 0.016).Mild caloric restriction, an accessible and affordable intervention, substantially reduced insulin doses in patients with T2D and SIR. Further studies are needed to determine if the intervention and results are sustainable outside of a hospital setting.

  14. Quarantine after an international biological weapons attack: medical and public health requirements for containment.

    Science.gov (United States)

    Oren, Meir

    2004-11-01

    The world now faces the dreadful possibility of biological weapons attacks by terrorists. Healthcare systems would have to cope with such emergencies should all preemptive measures fail. Information gained from the Global Mercury exercise and the SARS outbreak has shown that containing an outbreak at the start is more effective than reacting to it once it has spread and that containment should be treated both nationally and internationally. On the national level this entails developing rapid and effective methods to detect and identify infected cases, and implementing isolation and control measures to lower the risk of further transmission of the disease while assuring the safety of medical teams and laboratory workers. Strategic contingency plans should incorporate well-defined procedures for hospitalization and isolation of patients, providing regional backup of medical personnel and equipment and maintaining close cooperation between the various bodies in the healthcare system. Quarantine is an effective containment measure, especially if voluntarily imposed. Modern communication systems can help by sending professional teams timely instructions and providing the public with information to reduce panic and stress during quarantine procedures. Informing the public poses a dilemma: finding a balance between giving advance warning of an imminent epidemic outbreak and ascertaining the likelihood of its occurrence. Containment of international bioterrorist attacks depends entirely on close international cooperation to implement national and international strategic contingency plans with free exchange of information and recognition of procedures.

  15. French regulatory requirements concerning severe accidents in PWRs and associated research programme

    International Nuclear Information System (INIS)

    L'Homme, A.; Pelce, J.

    1983-12-01

    The French approach to safety doctrine is first presented: safety objectives as regards populations, and, plant safety objectives. Then, a description of ultimate or ''U'' procedures, involving the proceeding of physical phenomena induced by severe accidents, is given. Finally, R and D programs in relation to the various stages or severe accidents are presented

  16. French regulatory requirements concerning severe accidents in PWRs and associated research programme

    International Nuclear Information System (INIS)

    L'Homme, A.; Pelce, J.

    1986-07-01

    This report gives a global view of the French reactor safety approach; aspects in relation with severe accidents are pointed out: safety goals regarding population, and safety goals regarding plant design. Ultimate or U procedures involving physical phenomena of severe accidents are then described. R. and D. programs have been defined with regard to the priorities resulting from this approach [fr

  17. Main features of licensing requirements for nuclear installations in several OECD member countries

    International Nuclear Information System (INIS)

    Reyners, P.

    1977-01-01

    The present paper contains a brief description of the main features of the above-mentioned six countries' licensing systems, namely the legal regime applicable, the appropriate licensing bodies, the general frame and scope of the respective national regimes, the involvement of the public and technical safety bodies as well as the inspection procedures. This description is supplemented by some introductory remarks. (orig.) [de

  18. Main features of licensing requirements for nuclear installations in several OECD member countries

    International Nuclear Information System (INIS)

    Reyners, P.

    1975-01-01

    The present paper contains a brief description of the main features of the above-mentioned six countries' licensing systems, namely the legal regime applicable, the appropriate licensing bodies, the general frame and scope of the respective national regimes, the involvement of the public and technical safety bodies as well as the inspection procedures. This description is supplemented by some introductory remarks. (orig.) [de

  19. 43 CFR 404.60 - Does this rule contain an information collection that requires approval by the Office of...

    Science.gov (United States)

    2010-10-01

    ... 43 Public Lands: Interior 1 2010-10-01 2010-10-01 false Does this rule contain an information collection that requires approval by the Office of Management and Budget (OMB)? 404.60 Section 404.60 Public Lands: Interior Regulations Relating to Public Lands BUREAU OF RECLAMATION, DEPARTMENT OF THE INTERIOR RECLAMATION RURAL WATER SUPPLY PROGRAM...

  20. 77 FR 73294 - Requirements for Child-Resistant Packaging: Products Containing Imidazolines Equivalent to 0.08...

    Science.gov (United States)

    2012-12-10

    ... injury or illness than the PPPA standard; and (2) the state or political subdivision applies to the... for Child-Resistant Packaging: Products Containing Imidazolines Equivalent to 0.08 Milligrams or More... Commission (CPSC, Commission, or we) is issuing a rule to require child-resistant (CR) packaging for any over...

  1. 10 CFR 32.61 - Ice detection devices containing strontium-90; requirements for license to manufacture or...

    Science.gov (United States)

    2010-01-01

    ...; requirements for license to manufacture or initially transfer. 32.61 Section 32.61 Energy NUCLEAR REGULATORY COMMISSION SPECIFIC DOMESTIC LICENSES TO MANUFACTURE OR TRANSFER CERTAIN ITEMS CONTAINING BYPRODUCT MATERIAL... manufacture or initially transfer. An application for a specific license to manufacture or initially transfer...

  2. Interface requirements to couple thermal hydraulics codes to severe accident codes: ICARE/CATHARE

    Energy Technology Data Exchange (ETDEWEB)

    Camous, F.; Jacq, F.; Chatelard, P. [IPSN/DRS/SEMAR CE-Cadarache, St Paul Lez Durance (France)] [and others

    1997-07-01

    In order to describe with the same code the whole sequence of severe LWR accidents, up to the vessel failure, the Institute of Protection and Nuclear Safety has performed a coupling of the severe accident code ICARE2 to the thermalhydraulics code CATHARE2. The resulting code, ICARE/CATHARE, is designed to be as pertinent as possible in all the phases of the accident. This paper is mainly devoted to the description of the ICARE2-CATHARE2 coupling.

  3. Assessment of fission product release from the reactor containment building during severe core damage accidents in a PWR

    International Nuclear Information System (INIS)

    Fermandjian, J.; Evrard, J.M.; Generino, G.

    1984-07-01

    Fission product releases from the RCB associated with hypothetical core-melt accidents ABβ, S 2 CDβ and TLBβ in a PWR-900 MWe have been performed using French computer codes (in particular, the JERICHO Code for containment response analysis and AEROSOLS/B1 for aerosol behavior in the containment) related to thermalhydraulics and fission product behavior in the primary system and in the reactor containment building

  4. 78 FR 41298 - Children's Products Containing Lead; Procedures and Requirements for Exclusions From Lead Limits...

    Science.gov (United States)

    2013-07-10

    ... on materials previously submitted in connection with a petition for exclusion under this section. In... 16 CFR 1500.90 to provide procedures and requirements for evaluating products or materials for... and comment rulemaking, section 553 of the APA provides an exception when the agency, for good cause...

  5. Use of open source software in estimating the effects of a severe accident on the Mark II containment; Uso de software de fuente abierta en la estimacion de los efectos de un accidente severo sobre la contencion Mark II

    Energy Technology Data Exchange (ETDEWEB)

    Sainz, E.; Arguelles, R., E-mail: eduardo.sainz@inin.gob.mx [ININ, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico)

    2015-09-15

    Because the spectrum of scenarios of severe accident before which must verify the integrity of the containment can be very broad, it arises here a calculation methodology to estimate the structural response of the containment without incurring in high costs for commercial software licenses, or in times and calculation excessive requirements. The capabilities of computer programs with license of open source, OpenFOAM for CFD calculations and Salome-Meca for thermal and mechanical calculations were tested. The methodology begins of the venting of mass and energy that are postulated inside the container and the values of the thermal and mechanical fields are obtained through the walls. (Author)

  6. From Requirements via Colored Workflow Nets to an Implementation in Several Workflow Systems

    DEFF Research Database (Denmark)

    Mans, Ronny S.; van der Aalst, Willibrordus Martinus Pancratius; Molemann, A.J.

    2007-01-01

    Care organizations, such as hospitals, need to support complex and dynamic workflows. More- over, many disciplines are involved. This makes it important to avoid the typical disconnect between requirements and the actual implementation of the system. This paper proposes an approach where an Execu......Care organizations, such as hospitals, need to support complex and dynamic workflows. More- over, many disciplines are involved. This makes it important to avoid the typical disconnect between requirements and the actual implementation of the system. This paper proposes an approach where...... an Executable Use Case (EUC) and Colored Care organizations, such as hospitals, need to support complex and dynamic workflows. Moreover, many disciplines are involved. This makes it important to avoid the typical disconnect between requirements and the actual implementation of the system. This paper proposes...

  7. From Requirements via Colored Workflow Nets to an Implementation in Several Workflow Systems

    DEFF Research Database (Denmark)

    Mans, Ronnie S:; van der Aalst, Wil M.P.; Bakker, Piet J.M.

    2007-01-01

    care process of the Academic Medical Center (AMC) hospital is used as reference process. The process consists of hundreds of activities. These have been modeled and analyzed using an EUC and a CWN. Moreover, based on the CWN, the process has been implemented using four different workflow systems......Care organizations, such as hospitals, need to support complex and dynamic workflows. More- over, many disciplines are involved. This makes it important to avoid the typical disconnect between requirements and the actual implementation of the system. This paper proposes an approach where...... an Executable Use Case (EUC) and Colored Workflow Net (CWN) are used to close the gap between the given requirements specification and the realization of these requirements with the help of a workflow system. This paper describes a large case study where the diagnostic tra jectory of the gynaecological oncology...

  8. Input data requirements for special processors in the computation system containing the VENTURE neutronics code

    International Nuclear Information System (INIS)

    Vondy, D.R.; Fowler, T.B.; Cunningham, G.W.

    1979-07-01

    User input data requirements are presented for certain special processors in a nuclear reactor computation system. These processors generally read data in formatted form and generate binary interface data files. Some data processing is done to convert from the user oriented form to the interface file forms. The VENTURE diffusion theory neutronics code and other computation modules in this system use the interface data files which are generated

  9. Input data requirements for special processors in the computation system containing the VENTURE neutronics code

    International Nuclear Information System (INIS)

    Vondy, D.R.; Fowler, T.B.; Cunningham, G.W.

    1976-11-01

    This report presents user input data requirements for certain special processors in a nuclear reactor computation system. These processors generally read data in formatted form and generate binary interface data files. Some data processing is done to convert from the user-oriented form to the interface file forms. The VENTURE diffusion theory neutronics code and other computation modules in this system use the interface data files which are generated

  10. The technical requirements concerning severe accident management in nuclear power plants

    International Nuclear Information System (INIS)

    Okamoto, Koji; Sugiyama, Tomoyuki; Kamata, Shinya

    2014-01-01

    The Great East Japan Earthquake with a magnitude of 9.0 (The 2011 off the Pacific coast of Tohoku Earthquake) occurred on March 11, 2011, and the beyond design-basis tsunami descended on the Fukushima Daiichi Nuclear Power Plant by the earthquake. Eventually, the core cooling systems of the units 1, 2 and 3 could not operate stably, they all suffered severe accident, and hydrogen explosions were triggered in the reactor buildings of units 1, 3 and 4. In the light of these circumstances, Atomic Energy Society of Japan (AESJ) decided to establish a standard that consolidates the concept of maintaining and improving severe accident management. In the SAM standard, the combination of hardware and software measures based on the risk assessment enables a scientific and rational approach to apply to scenarios of various severe accidents including low-frequency, high-impact events, and assures safety with functionality and flexibility. The SAM standard is already established in March, 2014. After publication of the SAM standard, with regard to effectiveness assessment for accident management and treatment of the uncertainty of severe accident analysis code, for example, the detailed guideline will be prepared as appendices of the standard. (author)

  11. Bronchial thermoplasty: a novel treatment for severe asthma requiring monitored anesthesia care.

    Science.gov (United States)

    Lee, Jamille A; Rowen, David W; Rose, David D

    2011-12-01

    Dexmedetomidine used in monitored anesthesia care produces a safe and effective technique well documented in research. We report the successful use of dexmedetomidine for sedation during bronchial thermoplasty, a new treatment for patients with severe persistent asthma refractory to inhaled corticosteroids and long-term beta-2 agonists.

  12. [Neonatal ABO incompatibility underlies a potentially severe hemolytic disease of the newborn and requires adequate care].

    Science.gov (United States)

    Senterre, T; Minon, J-M; Rigo, J

    2011-03-01

    ABO allo-immunization is the most frequent hemolytic disease of the newborn and ABO incompatibility is present in 15-25 % of pregnancies. True ABO alloimmunization occurs in approximately one out of 150 births. Intensity is generally lower than in RhD allo-immunization. We report on three cases showing that ABO allo-immunization can lead to severe hemolytic disease of the newborn with potentially threatening hyperbilirubinemia and complications. Early diagnosis and adequate care are necessary to prevent complications in ABO incompatibility. A direct antiglobulin test is the cornerstone of diagnosis and should be performed at birth on cord blood sampling in all group infants born to O mothers, especially if of African origin. Risk factor analysis and attentive clinical monitoring during the first days of life are essential. Vigilance is even more important for infants discharged before the age of 72 h. Every newborn should be assessed for the risk of developing severe hyperbilirubinemia and should be examined by a qualified healthcare professional in the first days of life. Treatment depends on the total serum bilirubin level, which may increase very rapidly in the first 48 h of life in cases of hemolytic disease of the newborn. Phototherapy and, in severe cases, exchange transfusion are used to prevent hyperbilirubinemia encephalopathy. Intravenous immunoglobulins are used to reduce exchange transfusion. Treatments of severe hemolytic disease of the newborn should be provided and performed by trained personnel in neonatal intensive care units. Copyright © 2010 Elsevier Masson SAS. All rights reserved.

  13. Greater-than-Class C low-level radioactive waste shipping package/container identification and requirements study

    International Nuclear Information System (INIS)

    Tyacke, M.

    1993-08-01

    This report identifies a variety of shipping packages (also referred to as casks) and waste containers currently available or being developed that could be used for greater-than-Class C (GTCC) low-level waste (LLW). Since GTCC LLW varies greatly in size, shape, and activity levels, the casks and waste containers that could be used range in size from small, to accommodate a single sealed radiation source, to very large-capacity casks/canisters used to transport or dry-store highly radioactive spent fuel. In some cases, the waste containers may serve directly as shipping packages, while in other cases, the containers would need to be placed in a transport cask. For the purpose of this report, it is assumed that the generator is responsible for transporting the waste to a Department of Energy (DOE) storage, treatment, or disposal facility. Unless DOE establishes specific acceptance criteria, the receiving facility would need the capability to accept any of the casks and waste containers identified in this report. In identifying potential casks and waste containers, no consideration was given to their adequacy relative to handling, storage, treatment, and disposal. Those considerations must be addressed separately as the capabilities of the receiving facility and the handling requirements and operations are better understood

  14. Strategy of severe accident physical modeling in view of recent requirements to safety analysis

    International Nuclear Information System (INIS)

    Bolshov, L.A.

    1994-01-01

    Nuclear power destiny in various states including Russia is not free from questions. Where there is plenty of non-expensive natural gas or coal in a country, the competition of nuclear power with other power sources is especially intense. Until one considers the economic efficiency or environmental impact of the normally operating plant, the estimate of the proponents favorite choice may be rather optimistic in many cases. As soon as safety aspects of nuclear power are concerned it is necessary to answer very significant questions about the dangers resulting from severe accidents. TMI and, to a greater extent, Chernobyl, demonstrated the other aspect of the severe accident problem. It serves no purpose to dwell upon the inadequate reaction of the population on the radiation problem. It is of little use to try to prove that the health consequences of the Chernobyl or some other radiation accident are substantially overestimated. To make an advance one must substantially reduce the severe accident risk. Besides that is is necessary to give a convincing proof that such a reduction has really been made

  15. The TIR-domain containing adaptor TRAM is required for TLR7 mediated RANTES production.

    Directory of Open Access Journals (Sweden)

    Enda Shevlin

    Full Text Available Toll-like receptor 7 (TLR7 plays a vital role in the immune response to ssRNA viruses such as human rhinovirus (HRV and Influenza, against which there are currently no treatments or vaccines with long term efficacy available. Clearly, a more comprehensive understanding of the TLR7 signaling axis will contribute to its molecular targeting. TRIF related adaptor molecule (TRAM plays a vital role in TLR4 signaling by recruiting TRIF to TLR4, followed by endosomal trafficking of the complex and initiation of IRF3 dependent type I interferon production as well as NF-κB dependent pro-inflammatory cytokine production. Towards understanding the molecular mechanisms that regulate TLR7 functionality, we found that TRAM(-/- murine macrophages exhibited a transcriptional and translational impairment in TLR7 mediated RANTES, but not TNFα, production. Suppression of TRAM expression in human macrophages also resulted in an impairment in TLR7 mediated CCL5 and IFN-β, but not TNFα, gene induction. Furthermore, suppression of endogenous human TRAM expression in human macrophages significantly impaired RV16 induced CCL5 and IFNβ, but not TNFα gene induction. Additionally, TRAM-G2A dose-dependently inhibited TLR7 mediated activation of CCL5, IFNβ and IFNα reporter genes. TLR7-mediated phosphorylation and nuclear translocation of IRF3 was impaired in TRAM(-/- cells. Finally, co-immunoprecipitation studies indicated that TRAM physically interacts with MyD88 upon TLR7 stimulation, but not under basal conditions. Our results clearly demonstrate that TRAM plays a, hitherto unappreciated, role in TLR7 signaling through a novel signaling axis containing, but not limited to, MyD88, TRAM and IRF3 towards the activation of anti-viral immunity.

  16. WWER 440/213 NPP containment from the point of view of IAEA requirements and current European practice

    International Nuclear Information System (INIS)

    Sabata, M.

    2000-01-01

    In principle, in a NPP three barriers are used to prevent the release of radioactive substances into the environment: the fuel cladding, the primary circuit boundary, and the containment. The presentation deals with the third barrier - the containment, and explains the philosophy of maximum design accident management in the containment of WWER-440/213 NPPs. This type of containment is shown to be an original and fully functional technical solution. Due to the use of the large reserve of the H 3 BO 3 solution and to the active spray systems, an underpressure can be quickly established, thus minimizing any impact of the accident on the environment. The capability of the bubble condenser system for a maximum design accident has been proven by analyses and by tests performed within the framework of PHARE PH 2.13/95: the condenser system can reduce the pressure in the containment in an effective way. The majority of questions arising so far have been answered and the concern resulting from insufficient verification of the system with regard to western standards has been refuted. The WWER-440/213 is fully consistent with IAEA requirements and with current European practice. (A.K.)

  17. Severe Chronic Upper Airway Disease (SCUAD) in children. Definition issues and requirements.

    Science.gov (United States)

    Karatzanis, A; Kalogjera, L; Scadding, G; Velegrakis, S; Kawauchi, H; Cingi, C; Prokopakis, E

    2015-07-01

    Upper airway diseases are extremely common, and a significant proportion of patients are not adequately controlled by contemporary treatment algorithms. The term SCUAD (Severe Chronic Upper Airway Disease) has been previously introduced to describe such cases. However, this term has not been adequately focused on children. This study aims to address the necessity of the term, as well as further details specifically for children. For this purpose, a review was performed of the current literature, with specific focus on issues regarding SCUAD in children. Paediatric SCUAD represents a heterogeneous group of patients and has significant clinical and socioeconomic implications. Relevant literature is generally lacking and questions regarding definition and pathogenesis remain unanswered. Accurate definition and acknowledgement of paediatric SCUAD cases may lead to better design of future clinical and molecular research protocols. This may provide improved understanding of the underlying disease processes, more accurate data regarding socioeconomic burden, and, above all, more successful treatment and prevention strategies. Copyright © 2015 Elsevier Ireland Ltd. All rights reserved.

  18. Structural study of CH4, CO2 and H2O clusters containing from several tens to several thousands of molecules

    Science.gov (United States)

    Torchet, G.; Farges, J.; de Feraudy, M. F.; Raoult, B.

    Clusters are produced during the free jet expansion of gaseous CH4, CO2 or H2O. For a given stagnation temperature To, the mean cluster size is easily increased by increasing the stagnation pressure p0. On the other hand, the cluster temperature does not depend on stagnation conditions but mainly on properties of the condensed gas. An electron diffraction analysis provides information about the cluster structure. Depending on whether the diffraction patterns exhibit crystalline lines or not, the structure is worked out either by using crystallographic methods or by constructing cluster models. When they contain more than a few thousand molecules, clusters show a crystalline structure identical to that of one phase, namely, the cubic phase, known in bulk solid: plastic phase (CH4), unique solid phase (CO2) or metastable cubic phase (H2O). When decreasing the cluster size, the studied compounds behave quite differently: CO2 clusters keep the same crystalline structure, CH4 clusters show the multilayer icosahedral structure wich has been found in rare gas clusters, and H2O clusters adopt a disordered structure different from the amorphous structures of bulk ice. Des agrégats sont produits au cours de la détente en jet libre des gaz CH4, CO2 ou H2O. Pour une température initiale donnée To, on accroît facilement la taille moyenne des agrégats en augmentant la pression initiale po . Par contre, la température des agrégats dépend principalement des propriétés du gaz condensé. Une analyse par diffraction électronique permet l'étude de la structure des agrégats. Selon que les diagrammes de diffraction contiennent ou non des raies cristallines, on a recours soit à des méthodes cristallographiques soit à la construction de modèles d'agrégats. Lorsqu'ils renferment plus de quelques milliers de molécules, les agrégats adoptent la structure cristalline de l'une des phases connues du solide massif et plus précisément la phase cubique : phase plastique pour

  19. Some Examples of the Relationship Between Containment and Other Engineered Safeguard Requirements, Accident Analyses and Site Conditions

    Energy Technology Data Exchange (ETDEWEB)

    Vinck, W. F.; Maurer, H. [EURATOM, Brussels (Belgium)

    1967-09-15

    The paper refers primarily to nuclear power reactors for which EURATOM has performed safety reviews in co-operation with national technical advisory organizations concerned in the licensing procedures. Comparative data are tabulated on a number of containment concepts and other engineered safeguards to protect against or to limit the consequences of major hypothetical accidents for a number of power reactors. Main environmental data, such as magnitudes of exclusion areas and population densities, are also presented. A number of topics of particular interest which were encountered during the safety analysis and which find widespread application in the assessment of the siting conditions and emergency planning are discussed. In this discussion, emphasis is placed on containment, engineered safeguards and emergency equipment. These items are considered in general with emphasis on: (a ) the importance of meeting operability and high efficiency requirements (reliability ) when needed through design and layout; (b) periodic testing and/or inspection possibilities and requirements in order to maintain high availability standards. Attention is drawn to some difficulties which have arisen in connection with design, material choice and construction of steel containment structures and which in the interests of safety as well as of economic optimization justify in the future more care in the use of possibly uniform or single-code requirements. Examples of uncertainties encountered in some accident analyses and their influence on siting considerations are discussed, with emphasis on safeguards intended to retain radioactive material in the plant, such as extent of core damage, iodine plate-out, filtering efficiency, wash-out effects. The means by which some of the main uncertainties have been or can in the future be eliminated through appropriate experimental programmes performed throughout the world are discussed. Some examples are also given of the influence of factors which

  20. Comparison of computer codes relative to the aerosol behavior in the reactor containment building during severe core damage accidents in a PWR

    International Nuclear Information System (INIS)

    Fermandjian, J.; Bunz, H.; Dunbar, I.; Gauvain, J.; Ricchena, R.

    1986-01-01

    The present study concerns a comparative exercise, performed within the framework of the Commission of the European Communities, of the computer codes (AEROSIM-M, UK; AEROSOLS/B1, France; CORRAL-2, CEC and NAUA Mod5, Germany) used in order to assess the aerosol behavior in the reactor containment building during severe core damage accidents in a PWR. Topics considered in this paper include aerosols, containment buildings, reactor safety, fission product release, reactor cores, meltdown, and monitoring

  1. Effect of Spray System on Fission Product Distribution in Containment During a Severe Accident in a Two-Loop Pressurized Water Reactor

    Directory of Open Access Journals (Sweden)

    Mehdi Dehjourian

    2016-08-01

    Full Text Available The containment response during the first 24 hours of a low-pressure severe accident scenario in a nuclear power plant with a two-loop Westinghouse-type pressurized water reactor was simulated with the CONTAIN 2.0 computer code. The accident considered in this study is a large-break loss-of-coolant accident, which is not successfully mitigated by the action of safety systems. The analysis includes pressure and temperature responses, as well as investigation into the influence of spray on the retention of fission products and the prevention of hydrogen combustion in the containment.

  2. Compilation of Requirements for Safe Handling of Fluorine and Fluorine-Containing Products of Uranium Hexafluoride Conversion

    Energy Technology Data Exchange (ETDEWEB)

    Ferrada, J.J.

    2000-04-03

    Public Law (PL) 105-204 requires the U.S. Department of Energy to develop a plan for inclusion in the fiscal year 2000 budget for conversion of the Department's stockpile of depleted uranium hexafluoride (DUF{sub 6}) to a more stable form over an extended period. The conversion process into a more stable form will produce fluorine compounds (e.g., elemental fluorine or hydrofluoric acid) that need to be handled safely. This document compiles the requirements necessary to handle these materials within health and safety standards, which may apply in order to ensure protection of the environment and the safety and health of workers and the public. Fluorine is a pale-yellow gas with a pungent, irritating odor. It is the most reactive nonmetal and will react vigorously with most oxidizable substances at room temperature, frequently with ignition. Fluorine is a severe irritant of the eyes, mucous membranes, skin, and lungs. In humans, the inhalation of high concentrations causes laryngeal spasm and broncospasms, followed by the delayed onset of pulmonary edema. At sublethal levels, severe local irritation and laryngeal spasm will preclude voluntary exposure to high concentrations, unless the individual is trapped or incapacitated. A blast of fluorine gas on the shaved skin of a rabbit causes a second degree burn. Lower concentrations cause severe burns of insidious onset, resulting in ulceration, similar to the effects produced by hydrogen fluoride. Hydrofluoric acid is a colorless, fuming liquid or gas with a pungent odor. It is soluble in water with release of heat. Ingestion of an estimated 1.5 grams produced sudden death without gross pathological damage. Repeated ingestion of small amounts resulted in moderately advanced hardening of the bones. Contact of skin with anhydrous liquid produces severe burns. Inhalation of AHA or aqueous hydrofluoric acid mist or vapors can cause severe respiratory tract irritation that may be fatal. Based on the extreme chemical

  3. Structure and optical properties of several organic-inorganic hybrids containing corner-sharing chains of bismuth iodide octahedra.

    Science.gov (United States)

    Mitzi, D B; Brock, P

    2001-04-23

    Two organic-inorganic bismuth iodides of the form (H3N-R-NH3)BiI5 are reported, each containing long and relatively flexible organic groups, R. The norganic framework in each case consists of distorted BiI6 octahedra sharing cis vertexes to form zigzag chains. Crystals of (H3NC18H24S2NH3)BiI5 were grown from a slowly cooled ethylene glycol/2-butanol solution containing bismuth(III) iodide and AETH.2HI, where AETH = 1,6-bis[5'-(2' '-aminoethyl)-2'-thienyl]hexane. The new compound, (H2AETH)BiI5, adopts an orthorhombic (Aba2) cell with the lattice parameters a = 20.427(3) A, b = 35.078(5) A, c = 8.559(1) A, and Z = 8. The structure consists of corrugated layers of BiI5(2-) chains, with Bi-I bond lengths ranging from 2.942(3) to 3.233(3) A, separated by layers of the organic (H2AETH)(2+) cations. Crystals of the analogous (H3NC12H24NH3)BiI5 compound were also prepared from a concentrated aqueous hydriodic acid solution containing bismuth(III) iodide and the 1,12-dodecanediamine (DDDA) salt, DDDA.2HI. (H2DDDA)BiI5 crystallizes in an orthorhombic (Ibam) cell with a = 17.226(2) A, b = 34.277(4) A, c = 8.654(1) A, and Z = 8. The Bi-I bonds range in length from 2.929(1) to 3.271(1) A. While the inorganic chain structure is nearly identical for the two title compounds, as well as for the previously reported (H3NC6H12NH3)BiI5 [i.e., (H2DAH)BiI5] structure, the packing of the chains is strongly influenced by the choice of organic cation. Optical absorption spectra for thermally ablated thin films of the three organic-inorganic hybrids containing BiI5(2-) chains are reported as a function of temperature (25-290 K). The dominant long-wavelength feature in each case is attributed to an exciton band, which is apparent at room temperature and, despite the similar inorganic chain structure, varies in position from 491 to 541 nm (at 25 K).

  4. A simple and fast determination of microgram thorium in organic solution containing several hundreds times amount of uranium

    International Nuclear Information System (INIS)

    Yin Duanzhi; Cao Benhong; Yang Jinfeng

    1991-01-01

    Using spectrophotometric method, microgram thorium in 30% TBP-kerosene system containing large amount of uranium was successfully determined after one-step back-extraction with hydrochloric acid. The recovery of thorium is more than 98%, and the separation factor α U/Th is over 1 x 10 3 . Being reliable, simple and fast, the recommended method has been used in the research on spent fuel reprocessing and is expected applicable to other neutral phosphate extraction systems such as TOPO and DMHMP

  5. Possible presence of common tyvelose-containing glycans in Trichinella L1 larvae and embryonated eggs of several nematodes

    Directory of Open Access Journals (Sweden)

    Dea-Ayuela M.A.

    2001-06-01

    Full Text Available A monoclonal antibody (mAb US4 recognising an epitope containing tyvelose within the T. spiralis L-1 muscle larvae (TSL-1 antigens was tested in western-blot against various antigenic preparations from different stages of the following nematodes: T. spiralis (L1,adult, T. muris (egg, L1, L3, adult, Ascaris suum (egg, adult, Toxocara canis (egg, adult, Anisakis simplex (L3 and Haemochus contortus (egg. Positive reaction was present in antigen preparations from L1 larvae of T. spiralis and T. muris and from embryonated eggs of T. muris, A. suum, T. canis and H. conlortus.

  6. Perioperative analgesic requirements in severely obese adolescents and young adults undergoing laparoscopic versus robotic-assisted gastric sleeve resection

    Directory of Open Access Journals (Sweden)

    Anita Joselyn

    2015-01-01

    Full Text Available Purpose: One of the major advantages for patients undergoing minimally invasive surgery as compared to an open surgical procedure is the improved recovery profile and decreased opioid requirements in the perioperative period. There are no definitive studies comparing the analgesic requirements in patients undergoing two different types of minimally invasive procedure. This study retrospectively compares the perioperative analgesic requirements in severely obese adolescents and young adults undergoing laparoscopic versus robotic-assisted, laparoscopic gastric sleeve resection. Materials and Methods: With Institutional Review Board approval, the medication administration records of all severely obese patients who underwent gastric sleeve resection were retrospectively reviewed. Intra-operative analgesic and adjuvant medications administered, postoperative analgesic requirements, and visual analog pain scores were compared between those undergoing a laparoscopic procedure versus a robotic-assisted procedure. Results: This study cohort included a total of 28 patients who underwent gastric sleeve resection surgery with 14 patients in the laparoscopic group and 14 patients in the robotic-assisted group. Intra-operative adjuvant administration of both intravenous acetaminophen and ketorolac was similar in both groups. Patients in the robotic-assisted group required significantly less opioid during the intra-operative period as compared to patients in the laparoscopic group (0.15 ± 0.08 mg/kg vs. 0.19 ± 0.06 mg/kg morphine, P = 0.024. Cumulative opioid requirements for the first 72 postoperative h were similar in both the groups (0.64 ± 0.25 vs. 0.68 ± 0.27 mg/kg morphine, P = NS. No difference was noted in the postoperative pain scores. Conclusion: Although intraoperative opioid administration was lower in the robotic-assisted group, the postoperative opioid requirements, and the postoperative pain scores were similar in both groups.

  7. One adenosine deaminase allele in a patient with severe combined immunodeficiency contains a point mutation abolishing enzyme activity.

    OpenAIRE

    Valerio, D; Dekker, B M; Duyvesteyn, M G; van der Voorn, L; Berkvens, T M; van Ormondt, H; van der Eb, A J

    1986-01-01

    We have cloned and sequenced an adenosine deaminase (ADA) gene from a patient with severe combined immunodeficiency (SCID) caused by inherited ADA deficiency. Two point mutations were found, resulting in amino acid substitutions at positions 80 (Lys to Arg) and 304 (Leu to Arg) of the protein. Hybridization experiments with synthetic oligonucleotide probes showed that the determined mutations are present in both DNA and RNA from the ADA-SCID patient. In addition, wild-type sequences could be ...

  8. Chronic mould exposure as a risk factor for severe community acquired pneumonia in a patient requiring extra corporeal membrane oxygenation

    Directory of Open Access Journals (Sweden)

    Stephanie Thomas

    2015-01-01

    Full Text Available A previously fit and well man developed acute respiratory failure due to environmental mould exposure from living in damp rental accommodation. Despite aggressive intensive care management he rapidly deteriorated and required respiratory and cardiac Extracorporeal Membrane Oxygenation. We hypothesize that poor domiciliary conditions may make an underestimated contribution to community respiratory disease. These conditions may present as acute and severe illness with non-typical pathogens identified.

  9. Contain calculations of debris conditions adjacent to the BWR Mark I drywell shell during the later phases of a severe accident

    International Nuclear Information System (INIS)

    Hyman, C.R.

    1988-01-01

    Best estimate CONTAIN calculations have recently been performed by the BWR Severe Accident Technology (BWRSAT) Program at Oak Ridge National Laboratory to predict the primary containment response during the later phases of an unmitigated low-pressure Short Term Station Blackout at the Peach Bottom Atomic Power Station. Debris pour conditions leaving the failed reactor vessel are taken from the results of best estimate BWRSAR analyses that are based upon an assumed metallic debris melting temperature of 2750/degree/F (1783 K) and an oxide debris melting temperature of 4350/degree/F (2672 K). Results of the CONTAIN analysis for the case without sprays indicate failure of the drywell seals due to the extremely hot atmospheric conditions extant in the drywell. The maximum calculated temperature of the debris adjacent to the drywell shell is less than the melting temperature of the shell, yet the sustained temperatures may be sufficient to induce primary containment pressure boundary failure by the mechanism of creep-rupture. It is also predicted that a significant portion of the reactor pedestal wall is ablated during the period of the calculation. Nevertheless, the calculated results are recognized to be influenced by large modeling uncertainties. Several deficiencies in the application of the CORCON module within the CONTAIN code to BWR severe accident sequences are identified and discussed. 5 refs., 9 figs., 4 tabs.,

  10. Comparison of the thermal decomposition processes of several aminoalcohol-based ZnO inks with one containing ethanolamine

    Energy Technology Data Exchange (ETDEWEB)

    Gómez-Núñez, Alberto [University of Barcelona, Department of Electronics, Martí i Franquès 1, E08028-Barcelona (Spain); Roura, Pere [University of Girona, Department of Physics, Campus Montilivi, Edif. PII, E17071-Girona, Catalonia (Spain); López, Concepción [University of Barcelona, Department of Inorganic Chemistry, Martí i Franquès 1, E08028-Barcelona (Spain); Vilà, Anna, E-mail: avila@el.ub.edu [University of Barcelona, Department of Electronics, Martí i Franquès 1, E08028-Barcelona (Spain)

    2016-09-15

    Highlights: • Four alternatives to ethanolamine as stabilizer for the chemical synthesis of ZnO with zinc acetate dihydrate are proposed: aminopropanol, aminomethyl butanol, aminophenol and aminobenzyl alcohol. • Thermal decomposition processes described. Nitrogen cyclic compounds result. • Molecule flexibility helps decomposition, and in particular aliphatic aminoalcohols (quite flexible) decompose the precursor at lower temperatures than aromatic ones (more rigid). • Aminopropanol, aminomethyl butanol and aminobenzyl crystallize ZnO at a lower temperature than ethanolamine. • Nitrogen cyclic specimens have been identified and evolve in all cases (included ethanolamine) at temperatures up to 600 °C. - Abstract: Four inks for the production of ZnO semiconducting films have been prepared with zinc acetate dihydrate as precursor salt and one among the following aminoalcohols: aminopropanol (APr), aminomethyl butanol (AMB), aminophenol (APh) and aminobenzyl alcohol (AB) as stabilizing agent. Their thermal decomposition process has been analyzed in situ by thermogravimetric analysis (TGA), differential scanning calorimetry (DSC) and evolved gas analysis (EGA), whereas the solid product has been analysed ex-situ by X-ray diffraction (XRD) and infrared spectroscopy (IR). Although, except for the APh ink, crystalline ZnO is already obtained at 300 °C, the films contain an organic residue that evolves at higher temperature in the form of a large variety of nitrogen-containing cyclic compounds. The results indicate that APr can be a better stabilizing agent than ethanolamine (EA). It gives larger ZnO crystal sizes with similar carbon content. However, a common drawback of all the amino stabilizers (EA included) is that nitrogen atoms have not been completely removed from the ZnO film at the highest temperature of our experiments (600 °C).

  11. Experimental results from containment piping bellows subjected to severe accident conditions. Volume 1, Results from bellows tested in 'like-new' conditions

    International Nuclear Information System (INIS)

    Lambert, L.D.; Parks, M.B.

    1994-09-01

    Bellows are an integral part of the containment pressure boundary in nuclear power plants. They are used at piping penetrations to allow relative movement between piping and the containment wall, while minimizing the load imposed on the piping and wall. Piping bellows are primarily used in steel containments; however, they have received limited use in some concrete (reinforced and prestressed) containments. In a severe accident they may be subjected to pressure and temperature conditions that exceed the design values, along with a combination of axial and lateral deflections. A test program to determine the leak-tight capacity of containment penetration bellows is being conducted under the sponsorship of the US Nuclear Regulatory Commission at Sandia National Laboratories. Several different bellows geometries, representative of actual containment bellows, have been subjected to extreme deflections along with pressure and temperature loads. The bellows geometries and loading conditions are described along with the testing apparatus and procedures. A total of thirteen bellows have been tested, all in the 'like-new' condition. (Additional tests are planned of bellows that have been subjected to corrosion.) The tests showed that bellows are capable of withstanding relatively large deformations, up to, or near, the point of full compression or elongation, before developing leakage. The test data is presented and discussed

  12. Comparison of computer codes relative to the aerosol behavior in the reactor containment building during severe core damage accidents in a PWR

    International Nuclear Information System (INIS)

    Fermandjian, J.; Dunbar, I.; Gauvain, J.; Ricchena, R.

    1986-02-01

    The present study concerns a comparative exercise, performed within the framework of the Commission of the European Communities, of the computer codes (AEROSISM-M, UK; AEROSOLS/BI, France; CORRAL-2, CEC and NAUA Mod5, Germany) used in order to assess the aerosol behavior in the reactor containment building during severe core damage accidents in a PWR

  13. Biosafety and Biosecurity in European Containment Level 3 Laboratories: Focus on French Recent Progress and Essential Requirements

    Directory of Open Access Journals (Sweden)

    Boris Pastorino

    2017-05-01

    Full Text Available Even if European Union (EU Member States are obliged to implement EU Directives 2000/54/EC on the protection of workers from risks related to exposure to biological agents at work, national biosafety regulations and practices varied from country to country. In fact, EU legislation on biological agents and genetically modified microorganisms is often not specific enough to ensure harmonization leading to difficulties in implementation for most laboratories. In the same way, biosecurity is a relatively new concept and a few EU Member States are known to have introduced national laboratory biosecurity legislation. In France, recent regulations have reinforced biosafety/biosecurity in containment level 3 (CL-3 laboratories but they concern a specific list of pathogens with no correlation in other European Members States. The objective of this review was to summarize European biosafety/biosecurity measures concerning CL-3 facilities focusing on French specificities. Essential requirements needed to preserve efficient biosafety measures when manipulating risk group 3 biological agents are highlighted. In addition, International, European and French standards related to containment laboratory planning, operation or biosafety equipment are described to clarify optimal biosafety and biosecurity requirements.

  14. Biosafety and Biosecurity in European Containment Level 3 Laboratories: Focus on French Recent Progress and Essential Requirements.

    Science.gov (United States)

    Pastorino, Boris; de Lamballerie, Xavier; Charrel, Rémi

    2017-01-01

    Even if European Union (EU) Member States are obliged to implement EU Directives 2000/54/EC on the protection of workers from risks related to exposure to biological agents at work , national biosafety regulations and practices varied from country to country. In fact, EU legislation on biological agents and genetically modified microorganisms is often not specific enough to ensure harmonization leading to difficulties in implementation for most laboratories. In the same way, biosecurity is a relatively new concept and a few EU Member States are known to have introduced national laboratory biosecurity legislation. In France, recent regulations have reinforced biosafety/biosecurity in containment level 3 (CL-3) laboratories but they concern a specific list of pathogens with no correlation in other European Members States. The objective of this review was to summarize European biosafety/biosecurity measures concerning CL-3 facilities focusing on French specificities. Essential requirements needed to preserve efficient biosafety measures when manipulating risk group 3 biological agents are highlighted. In addition, International, European and French standards related to containment laboratory planning, operation or biosafety equipment are described to clarify optimal biosafety and biosecurity requirements.

  15. Collective radiation doses following a hypothetical, very severe accident to an irradiated fuel transport flask containing AGR fuel

    International Nuclear Information System (INIS)

    Corbett, J.O.

    1985-05-01

    Studies of the consequences of very severe, although unlikely, accidents to irradiated fuel transport flasks are made in order to evaluate risks. If an irradiated fuel transport flask carrying AGR fuel were damaged in a hypothetical accident involving a severe impact followed by a prolonged fire, a small proportion of caesium and other fission products might be released to the atmosphere from the gap inventory of broken fuel pins. The consequent radiation dose to the public would arise predominantly by direct irradiation from ground deposits and the ingestion of slightly contaminated foodstuffs. Although these collective doses must generally be estimated with the aid of computer codes, it is shown here that the worst case, when a high proportion of the radioactivity is deposited in a densely population area, can be assessed approximately by a much simpler method, an approach which is of great value in explaining the calculation in a manner that can be readily understood. A comparison is made between the simple approach and equivalent results from the NECTAR code, the worst case is compared with an ensemble average over all weather conditions, and the relative contributions of the two main routes to collective dose are discussed. (author)

  16. Analysis of radionuclide behavior in a BWR Mark-II containment under severe accident management condition in low pressure sequence

    International Nuclear Information System (INIS)

    Funayama, Kyoko; Kajimoto, Mitsuhiro; Nagayoshi, Takuji; Tanaka, Nobuo

    1999-01-01

    In the Level 2 PSA program at INS/NUPEC, MELCOR1.8.3 is extensively applied to analyze radionuclide behavior of dominant sequences. In addition, the revised source terms provided in the NUREG-1465 report have been also discussed to examine the potential of the radionuclides release to the environment in the conventional siting criteria. In the present study, characteristics of source terms to the environment were examined comparing with results by the Hypothetical Accident (LOCA), NUREG-1465 and MELCOR1.8.3. calculation for a typical BWR with a Mark-II containment in order to assure conservatives of the Hypothetical Accident in Japan. Release fractions of iodine to the environment for the Hypothetical Accident and NUREG-1465, which used engineering models for predicting radionuclide behaviors, were about 10 -4 and 10 -6 of core inventory, respectively, while the best estimate MELCOR1.8.3 code predicted 10 -9 of iodine to the environment. The present study showed that the engineering models in the Hypothetical Accident or NUREG-1465 have large conservatives to estimate source term of iodine to the environment. (author)

  17. Potential applicability of fuzzy set theory to analyses of containment response and uncertainty for postulated severe accidents

    International Nuclear Information System (INIS)

    Chun, M.H.; Ahn, K.I.

    1991-01-01

    An important issue faced by contemporary risk analysts of nuclear power plants is how to deal with uncertainties that arise in each phase of probabilistic risk assessments. The major uncertainty addressed in this paper is the one that arises in the accident-progression event trees (APETs), which treat the physical processes affecting the core after an initiating event occurs. Recent advances in the theory of fuzzy sets make it possible to analyze the uncertainty related to complex physical phenomena that may occur during a severe accident of nuclear power plants by means of fuzzy set or possibility concept. The main purpose of this paper is to prevent the results of assessment of the potential applicability of the fuzzy set theory to the uncertainty analysis of APETs as a possible alternative procedure to that used in the most recent risk assessment

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

    Energy Technology Data Exchange (ETDEWEB)

    Bakalov, Ivan [Gesellschaft fuer Anlagen- und Reaktorsicherheit (GRS) gGmbH, Berlin (Germany); Sonnenkalb, Martin [Gesellschaft fuer Anlagen- und Reaktorsicherheit (GRS) gGmbH, Koeln (Germany)

    2018-02-15

    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.

  19. 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.

  20. Mammalian target of rapamycin is required for phrenic long-term facilitation following severe but not moderate acute intermittent hypoxia.

    Science.gov (United States)

    Dougherty, Brendan J; Fields, Daryl P; Mitchell, Gordon S

    2015-09-01

    Phrenic long-term facilitation (pLTF) is a persistent increase in phrenic nerve activity after acute intermittent hypoxia (AIH). Distinct cell-signaling cascades give rise to pLTF depending on the severity of hypoxemia within hypoxic episodes. Moderate AIH (mAIH; three 5-min episodes, PaO2 ∼35-55 mmHG) elicits pLTF by a serotonin (5-HT)-dependent mechanism that requires new synthesis of brain-derived neurotrophic factor (BDNF), activation of its high-affinity receptor (TrkB), and ERK MAPK signaling. In contrast, severe AIH (sAIH; three 5-min episodes, PaO2 ∼25-30 mmHG) elicits pLTF by an adenosine-dependent mechanism that requires new TrkB synthesis and Akt signaling. Although both mechanisms require spinal protein synthesis, the newly synthesized proteins are distinct, as are the neurochemicals inducing plasticity (serotonin vs. adenosine). In many forms of neuroplasticity, new protein synthesis requires translational regulation via mammalian target of rapamycin (mTOR) signaling. Since Akt regulates mTOR activity, we hypothesized that mTOR activity is necessary for sAIH- but not mAIH-induced pLTF. Phrenic nerve activity in anesthetized, paralyzed, and ventilated rats was recorded before, during, and 60 min after mAIH or sAIH. Rats were pretreated with intrathecal injections of 20% DMSO (vehicle controls) or rapamycin (0.1 mM, 12 μl), a selective mTOR complex 1 inhibitor. Consistent with our hypothesis, rapamycin blocked sAIH- but not mAIH-induced pLTF. Thus spinal mTOR activity is required for adenosine-dependent (sAIH) but not serotonin-dependent (mAIH) pLTF, suggesting that distinct mechanisms regulate new protein synthesis in these forms of spinal neuroplasticity. Copyright © 2015 the American Physiological Society.

  1. Estradiol-induced increase in novel object recognition requires hippocampal NR2B-containing NMDA receptors.

    Science.gov (United States)

    Vedder, Lindsey C; Smith, Caroline C; Flannigan, Alaina E; McMahon, Lori L

    2013-01-01

    17β-estradiol (E2), at high circulating levels, enhances learning and memory in many women, making it a clinical treatment for hormone-related cognitive decline in aging. However, the mechanisms stimulated by E2, which are responsible for its cognitive enhancing effects, remain incompletely defined. Using an ovariectomized rat model, we previously reported that increasing plasma E2 enhances the magnitude of long-term potentiation (LTP) at hippocampal CA3-CA1 synapses, which is caused by a selective increase in current mediated by NR2B-containing NMDARs, leading to an increase in the NMDAR/AMPAR ratio. Whether the increase in NR2B current is causally related to the ability of E2 to enhance hippocampal dependent learning and memory has yet to be tested. Here, we find that E2 enhances performance in the novel object recognition (NOR) task with the same time course we previously showed E2 enhances the LTP magnitude, temporally linking the increase in LTP to enhanced learning and memory. Furthermore, using the selective NR2B subunit antagonist Ro25-6981, we find that the E2-enhanced NOR, like the enhanced LTP, requires hippocampal NR2B-containing NMDARs, specifically in area CA1. Finally, using whole-cell recordings and the phosphatase inhibitor orthovanadate, we investigated whether the E2-induced increase in NMDAR current is caused by an increase in the density of synaptic NMDARs and/or an increase in NMDAR subunit phosphorylation. We find that both mechanisms are responsible for the enhanced NMDAR current in E2-treated rats. Our results show that the E2-enhanced NOR requires a functional increase in NR2B-containing NMDARs, a requirement shared with the E2-enhanced LTP magnitude at CA3-CA1 synapses, supporting the hypothesis that the increase in LTP likely contributes to the enhanced learning and memory following an increase in plasma E2 levels. Copyright © 2012 Wiley Periodicals, Inc.

  2. A Novel Phytophthora sojae Resistance Rps12 Gene Mapped to a Genomic Region That Contains Several Rps Genes.

    Science.gov (United States)

    Sahoo, Dipak K; Abeysekara, Nilwala S; Cianzio, Silvia R; Robertson, Alison E; Bhattacharyya, Madan K

    2017-01-01

    Phytophthora sojae Kaufmann and Gerdemann, which causes Phytophthora root rot, is a widespread pathogen that limits soybean production worldwide. Development of Phytophthora resistant cultivars carrying Phytophthora resistance Rps genes is a cost-effective approach in controlling this disease. For this mapping study of a novel Rps gene, 290 recombinant inbred lines (RILs) (F7 families) were developed by crossing the P. sojae resistant cultivar PI399036 with the P. sojae susceptible AR2 line, and were phenotyped for responses to a mixture of three P. sojae isolates that overcome most of the known Rps genes. Of these 290 RILs, 130 were homozygous resistant, 12 heterzygous and segregating for Phytophthora resistance, and 148 were recessive homozygous and susceptible. From this population, 59 RILs homozygous for Phytophthora sojae resistance and 61 susceptible to a mixture of P. sojae isolates R17 and Val12-11 or P7074 that overcome resistance encoded by known Rps genes mapped to Chromosome 18 were selected for mapping novel Rps gene. A single gene accounted for the 1:1 segregation of resistance and susceptibility among the RILs. The gene encoding the Phytophthora resistance mapped to a 5.8 cM interval between the SSR markers BARCSOYSSR_18_1840 and Sat_064 located in the lower arm of Chromosome 18. The gene is mapped 2.2 cM proximal to the NBSRps4/6-like sequence that was reported to co-segregate with the Phytophthora resistance genes Rps4 and Rps6. The gene is mapped to a highly recombinogenic, gene-rich genomic region carrying several nucleotide binding site-leucine rich repeat (NBS-LRR)-like genes. We named this novel gene as Rps12, which is expected to be an invaluable resource in breeding soybeans for Phytophthora resistance.

  3. A Novel Phytophthora sojae Resistance Rps12 Gene Mapped to a Genomic Region That Contains Several Rps Genes.

    Directory of Open Access Journals (Sweden)

    Dipak K Sahoo

    Full Text Available Phytophthora sojae Kaufmann and Gerdemann, which causes Phytophthora root rot, is a widespread pathogen that limits soybean production worldwide. Development of Phytophthora resistant cultivars carrying Phytophthora resistance Rps genes is a cost-effective approach in controlling this disease. For this mapping study of a novel Rps gene, 290 recombinant inbred lines (RILs (F7 families were developed by crossing the P. sojae resistant cultivar PI399036 with the P. sojae susceptible AR2 line, and were phenotyped for responses to a mixture of three P. sojae isolates that overcome most of the known Rps genes. Of these 290 RILs, 130 were homozygous resistant, 12 heterzygous and segregating for Phytophthora resistance, and 148 were recessive homozygous and susceptible. From this population, 59 RILs homozygous for Phytophthora sojae resistance and 61 susceptible to a mixture of P. sojae isolates R17 and Val12-11 or P7074 that overcome resistance encoded by known Rps genes mapped to Chromosome 18 were selected for mapping novel Rps gene. A single gene accounted for the 1:1 segregation of resistance and susceptibility among the RILs. The gene encoding the Phytophthora resistance mapped to a 5.8 cM interval between the SSR markers BARCSOYSSR_18_1840 and Sat_064 located in the lower arm of Chromosome 18. The gene is mapped 2.2 cM proximal to the NBSRps4/6-like sequence that was reported to co-segregate with the Phytophthora resistance genes Rps4 and Rps6. The gene is mapped to a highly recombinogenic, gene-rich genomic region carrying several nucleotide binding site-leucine rich repeat (NBS-LRR-like genes. We named this novel gene as Rps12, which is expected to be an invaluable resource in breeding soybeans for Phytophthora resistance.

  4. Physiological Requirements to Perform the Glittre Activities of Daily Living Test by Subjects With Mild-to-Severe COPD.

    Science.gov (United States)

    Souza, Gérson F; Moreira, Graciane L; Tufanin, Andréa; Gazzotti, Mariana R; Castro, Antonio A; Jardim, José R; Nascimento, Oliver A

    2017-08-01

    The Glittre activities of daily living (ADL) test is supposed to evaluate the functional capacity of COPD patients. The physiological requirements of the test and the time taken to perform it by COPD patients in different disease stages are not well known. The objective of this work was to compare the metabolic, ventilatory, and cardiac requirements and the time taken to carry out the Glittre ADL test by COPD subjects with mild, moderate, and severe disease. Spirometry, Medical Research Council questionnaire, cardiopulmonary exercise test, and 2 Glittre ADL tests were evaluated in 62 COPD subjects. Oxygen uptake (V̇ O 2 ), carbon dioxide production, pulmonary ventilation, breathing frequency, heart rate, S pO 2 , and dyspnea were analyzed before and at the end of the tests. Maximum voluntary ventilation, Glittre peak V̇ O 2 /cardiopulmonary exercise test (CPET) peak V̇ O 2 , Glittre V̇ E /maximum voluntary ventilation, and Glittre peak heart rate/CPET peak heart rate ratios were calculated to analyze their reserves. Subjects carried out the Glittre ADL test with similar absolute metabolic, ventilatory, and cardiac requirements. Ventilatory reserve decreased progressively from mild to severe COPD subjects ( P reserve than the mild and moderate subjects ( P = .006 and P = .043, respectively) and significantly lower Glittre peak heart rate/CPET peak heart rate than mild subjects ( P = .01). Time taken to carry out the Glittre ADL test was similar among the groups ( P = .82 for GOLD 1 vs GOLD 2, P = .19 for GOLD 1 vs GOLD 3, and P = .45 for GOLD 2 vs GOLD 3). As the degree of air-flow obstruction progresses, the COPD subjects present significant lower ventilatory reserve to perform the Glittre ADL test. In addition, metabolic and cardiac reserves may differentiate the severe subjects. These variables may be better measures to differentiate functional performance than Glittre ADL time. Copyright © 2017 by Daedalus Enterprises.

  5. Incidence of Severe Osteonecrosis Requiring Total Joint Arthroplasty in Children and Young Adults Treated for Leukemia or Lymphoma

    DEFF Research Database (Denmark)

    Niinimäki, Riitta; Hansen, Lene Mølgaard; Niinimäki, Tuukka

    2013-01-01

    diagnosis codes given before the age of 40 were also retrieved. Results: The estimated cumulative incidence of TJA was 4.5% at 20 years for patients treated for chronic myeloid leukemia, followed by 2.1% for patients treated for acute myeloid leukemia. It was considerably lower in patients with acute...... the age of 10 (HR=24; 95% CI: 3.1-176 and HR=26; 95% CI: 3.6-192 respectively). Conclusion: The incidence of ON requiring TJA was highest among patients with myeloid leukemias and lowest in patients treated for ALL. Allo-SCT and age ≥10 years at diagnosis were the most important risk factors......Purpose: The population-based incidence of severe osteonecrosis (ON) necessitating total joint arthroplasty (TJA) in patients with hematological cancer is unknown. This study assessed the incidence of ON requiring primary TJA in children and young adults treated for leukemia or lymphoma. Methods...

  6. Vaccinia protein F12 has structural similarity to kinesin light chain and contains a motor binding motif required for virion export.

    Directory of Open Access Journals (Sweden)

    Gareth W Morgan

    2010-02-01

    Full Text Available Vaccinia virus (VACV uses microtubules for export of virions to the cell surface and this process requires the viral protein F12. Here we show that F12 has structural similarity to kinesin light chain (KLC, a subunit of the kinesin-1 motor that binds cargo. F12 and KLC share similar size, pI, hydropathy and cargo-binding tetratricopeptide repeats (TPRs. Moreover, molecular modeling of F12 TPRs upon the crystal structure of KLC2 TPRs showed a striking conservation of structure. We also identified multiple TPRs in VACV proteins E2 and A36. Data presented demonstrate that F12 is critical for recruitment of kinesin-1 to virions and that a conserved tryptophan and aspartic acid (WD motif, which is conserved in the kinesin-1-binding sequence (KBS of the neuronal protein calsyntenin/alcadein and several other cellular kinesin-1 binding proteins, is essential for kinesin-1 recruitment and virion transport. In contrast, mutation of WD motifs in protein A36 revealed they were not required for kinesin-1 recruitment or IEV transport. This report of a viral KLC-like protein containing a KBS that is conserved in several cellular proteins advances our understanding of how VACV recruits the kinesin motor to virions, and exemplifies how viruses use molecular mimicry of cellular components to their advantage.

  7. Comparison of european computer codes relative to the aerosol behavior in PWR containment buildings during severe core damage accidents. (Modelling of steam condensation on the particles)

    International Nuclear Information System (INIS)

    Bunz, H.; Dunbar, L.H.; Fermandjian, J.; Lhiaubet, G.

    1987-11-01

    An aerosol code comparison exercise was performed within the framework of the Commission of European Communities (Division of Safety of Nuclear Installations). This exercise, focused on the process of steam condensation onto the aerosols occurring in PWR containment buildings during severe core damage accidents, has allowed to understand the discrepancies between the results obtained. These discrepancies are due, in particular, to whether the curvature effect is modelled or not in the codes

  8. The Euratom-Rosatom ERCOSAM-SAMARA projects on containment thermal-hydraulics of current and future LWRs for severe accident management

    International Nuclear Information System (INIS)

    Paladino, D.; Guentay, S.; Andreani, M.; Tkatschenko, I.; Brinster, J.; Dabbene, F.; Kelm, S.; Allelein, H. J.; Visser, D. C.; Benz, S.; Jordan, T.; Liang, Z.; Porcheron, E.; Malet, J.; Bentaib, A.; Kiselev, A.; Yudina, T.; Filippov, A.; Khizbullin, A.; Kamnev, M.; Zaytsev, A.; Loukianov, A.

    2012-01-01

    During a postulated severe accident with core degradation, hydrogen would form in the reactor pressure vessel mainly due to high temperatures zirconium-steam reaction and flow together with steam into the containment where it will mix with the containment atmosphere (steam-air). The hydrogen transport into the containment is a safety concern because it can lead to explosive mixtures through the associated phenomena of condensation, mixing and stratification. The ERCOSAM and SAMARA projects, co-financed by the European Union and the Russia, include various experiments addressing accident scenarios scaled down from existing plant calculations to different thermal-hydraulics facilities (TOSQAN, MISTRA, PANDA, SPOT). The tests sequences aim to investigate hydrogen concentration build-up and stratification during a postulated accident and the effect of the activation of Severe Accident Management systems (SAMs), e.g. sprays, coolers and Passive Auto-catalytic Recombiners (PARs). Analytical activities, performed by the project participants, are an essential component of the projects, as they aim to improve and validate various computational methods. They accompany the projects in the various phases; plant calculations, scaling to generic containment and to the different facilities, planning pre-test and post-test simulations are performed. Code benchmark activities on the basis of conceptual near full scale HYMIX facility will finally provide a further opportunity to evaluate the applicability of the various methods to the study of scaling issues. (authors)

  9. Intermodal container transport logistics to andfrom Malaysian ports : Evaluation of Customer requirements and environmental eff‡ects

    OpenAIRE

    Nasir, Sharin

    2014-01-01

    Malaysian ports’ container volumes are expected to increase to 36.6 million TEUs in 2020 compare to 12 million TEUs in 2005. Almost 45% of the container volumes are local containers entering the Malaysian hinterland. The hinterland container transport movements are dominated by road haulage (90%), alongside road-rail intermodal that currently handles the remaining 10%. The aim of this research is to develop possible strategies for improving the logistics of the intermodal hinterland container...

  10. Psychometric study of the Required Care Levels for People with Severe Mental Disorder Assessment Scale (ENAR-TMG).

    Science.gov (United States)

    Lascorz, David; López, Victoria; Pinedo, Carmen; Trujols, Joan; Vegué, Joan; Pérez, Víctor

    2016-03-08

    People with severe mental disorder have significant difficulties in everyday life that involve the need for continued support. These needs are not easily measurable with the currently available tools. Therefore, a multidimensional scale that assesses the different levels of need for care is proposed, including a study of its psychometric properties. One-hundred and thirty-nine patients (58% men) with a severe mental disorder were assessed using the Required Care Levels for People with Severe Mental Disorder Assessment Scale (ENAR-TMG), the Camberwell Assessment of Need scale, and the Health of the Nation Outcome Scales. ENAR-TMG's psychometric features were examined by: a) evaluating 2 sources of validity evidence (evidence based on internal structure and evidence based on relations to other variables), and b) estimating the internal consistency, temporal stability, inter-rater reliability, and sensitivity to change of scores of the ENAR-TMG's subscales. Exploratory factor analyses revealed a one-factor structure for each of the theoretical dimensions of the scale, in which all but one showed a significant and positive correlation with the Camberwell Assessment of Need (range of r: 0.143-0.557) and Health of the Nation Outcome Scales (range of r: 0.241-0.474) scales. ENAR-TMG subscale scores showed acceptable internal consistency (range of ordinal α coefficients: 0.682-0.804), excellent test-retest (range of intraclass correlation coefficients: 0.889-0.999) and inter-rater reliabilities (range of intraclass correlation coefficients: 0.926-0.972), and satisfactory sensitivity to treatment-related changes (range of η 2 : 0.003-0.103). The satisfactory psychometric behaviour of the ENAR-TMG makes the scale a promising tool to assess global functioning in people with a severe mental disorder. Copyright © 2016 SEP y SEPB. Published by Elsevier España. All rights reserved.

  11. Comparison between MARCH-3 and MAAP-3 thermal-hydraulic results for a severe accident in a BWR system with MARK-III containment

    International Nuclear Information System (INIS)

    Barbucci, P.; Guidi, L.; Mariotti, G.

    1988-01-01

    A comparison between results provided by the Source Term Code Package and by the MAAP-3 code for a PWR with full pressure containment was presented. Thereafter the same two methodologies were used to analyse a severe accident sequence in a typical BWR power plant equipped with a General Electric BWR 6 reactor, rated at 2894 MWt, and a MARK-III type containment. As a reference sequence the TQUV was chosen. This sequence is characterized by a transient (T) with loss of feedwater (Q) and loss of all Emergency Core Cooling Systems, both at high pressure (U) and, after the intervention of the Automatic Depressurization System (ADS), at low pressure (V). After the vessel, failure two basic scenarios for the containment response were analysed: in the first one the pedestal is always dry, in the second one it is fully flooded. Typical limestone/common sand and basaltic concrete compositions were considered. In the following sections the obtained results will be shown with the main purpose to point out the different phenomenological models of the two codes rather than to look for the true plant response to such a severe accident. After the presentation of the most important physical models and of the main assumptions for the analyses (sects. 2 and 4), the comparison will be performed for the in-vessel phase, in section 3, and for the ex-vessel phase, in section 5

  12. Severe Corticosteroid-Induced Ocular Hypertension Requiring Bilateral Trabeculectomies in a Patient with Takayasu’s Arteritis

    Directory of Open Access Journals (Sweden)

    Anna Maria Gruener

    2016-01-01

    Full Text Available We present a rare case of severe corticosteroid-induced ocular hypertension (OHT after prolonged systemic corticosteroid use in a young woman with Takayasu’s arteritis. As she did not sufficiently respond to ocular antihypertensive therapies, bilateral enhanced trabeculectomies were required to normalize her intraocular pressures. The systemic side effects of corticosteroids are well known, yet steroid-induced OHT and glaucoma remain silent causes of ocular morbidity. This case highlights the importance of IOP-monitoring in visually asymptomatic patients on systemic corticosteroids. It further emphasizes the need to raise awareness of the potential ocular side effects of steroids amongst physicians, in particular those looking after patients with autoimmune and inflammatory diseases.

  13. Has the symptom severity inclusion requirement narrowed the definition of major depressive disorder in antidepressant efficacy trials?

    Science.gov (United States)

    Zimmerman, Mark; Walsh, Emily; Chelminski, Iwona; Dalrymple, Kristy

    2017-03-15

    The inclusion criteria of all placebo-controlled studies of antidepressants have required a minimum level of severity on standardized measures of symptoms of depression. In the present report from the Rhode Island Methods to Improve Diagnostic Assessment and Services (MIDAS) project we examined the association between scores on the Hamilton Depression Rating Scale (HAMD) and the number of criteria met for MDD, as well as the impact of different HAMD cutoff scores on the distribution of the number of DSM-IV criteria met. We speculated that the use of a minimum symptom severity score (MSSS) for inclusion in an antidepressant efficacy trial (AETs) disproportionately excludes patients who are at or just above the diagnostic threshold for MDD, whereas patients who are well above the diagnostic threshold are not excluded. Seven hundred forty outpatients with current MDD were evaluated with a semi-structured diagnostic interview. We compared the distribution of DSM-IV MDD criteria scores in patients who scored at or above or below the 3 cutoff scores on the HAMD most commonly used for inclusion in an AET. The distribution of the number of DSM-IV MDD symptom criteria met was significantly associated with HAMD scores. Compared to patients scoring below 18 on the HAMD the patients scoring 18 and above were less likely to report 5 MDD criteria (13.9% vs. 43.7%, χ 2 =82.2, pconducted in a single outpatient practice in which the majority of patients were white, female, and had health insurance. Although the study was limited to a single site, a strength of the recruitment procedure was that the sample was not selected for participation in a treatment study, and exclusion and inclusion criteria did not reduce the representativeness of the patient groups. While there is not a perfect relationship between the HAMD score and the number of DSM MDD criteria present, the results of the current study suggest that HAMD scores can be thought of as a proxy for the number of DSM

  14. Response of immunocompetent cells of bone marrow and spleen of mouse males of several strains to stress and to pyrazine containing chemosignals

    Directory of Open Access Journals (Sweden)

    Eugene V Daev

    2012-06-01

    Full Text Available The quantity of antibody producing cells and mitotic disturbances in dividing bone marrow cells of mice were studied after exposure of animals to a physical stressor or various pyrazinecontaining chemosignals. Several different strains of mice were used. We demonstrate that immune suppression and destabilization of the chromosome apparatus in dividing cells depend on: а mouse genotype and b side chains position  in the pyrazine ring. Importance of this effects in the light of wide usage of pyrazine containing substances in perfume industry, food production and pharmacology is discussed.

  15. CD4+ T cells are required to contain early extrathoracic TB dissemination and sustain multi-effector functions of CD8+ T and CD3− lymphocytes

    Science.gov (United States)

    Yao, Shuyu; Huang, Dan; Chen, Crystal Y.; Halliday, Lisa; Wang, Richard C.; Chen, Zheng W.

    2014-01-01

    The possibility that CD4+ T cells can act as “innate-like” cells to contain very-early M. tuberculosis (Mtb) dissemination and function as master helpers to sustain multiple effector functions of CD8+ T cells and CD3-negative lymphocytes during development of adaptive immunity against primary tuberculosis(TB) has not been demonstrated. We showed that pulmonary Mtb infection of CD4-depleted macaques surprisingly led to very-early extrathoracic Mtb dissemination, whereas CD4 deficiency clearly resulted in rapid TB progression. CD4 depletion during Mtb infection revealed the ability of CD8+ T cells to compensate and rapidly differentiate to Th17-like/Th1-like, and cytotoxic-like effectors, but these effector functions were subsequently unsustainable due to CD4 deficiency. While CD3-negative non-T lymphocytes in presence of CD4+ T cells developed predominant Th22-like and NK-like (perforin production) responses to Mtb infection, CD4 depletion abrogated these Th22-/NK-like effector functions and favored IL-17 production by CD3-negative lymphocytes. CD4-depleted macaques exhibited no or few pulmonary T effector cells constitutively producing IFN-γ, TNFα, IL-17, IL-22, and perforin at the endpoint of more severe TB, but presented pulmonary IL-4+ T effectors. TB granulomas in CD4-depleted macaques contained fewer IL-22+ and perforin+ cells despite presence of IL-17+ and IL-4+ cells. These results implicate previously-unknown “innate-like” ability of CD4+ T cells to contain extrathoracic Mtb dissemination at very early stage. Data also suggest that CD4+ T cells are required to sustain multiple effector functions of CD8+ T cells and CD3-negative lymphocytes and to prevent rapid TB progression during Mtb infection of nonhuman primates. PMID:24489088

  16. Therapy of Venezuelan patients with severe mucocutaneous or early lesions of diffuse cutaneous leishmaniasis with a vaccine containing pasteurized Leishmania promastigotes and bacillus Calmette-Guerin: preliminary report

    Directory of Open Access Journals (Sweden)

    Jacinto Convit

    2004-02-01

    Full Text Available Severe mucocutaneous (MCL and diffuse (DCL forms of American cutaneous leishmaniasis (ACL are infrequent in Venezuela. Chemotherapy produces only transitory remission in DCL, and occasional treatment failures are observed in MCL. We have evaluated therapy with an experimental vaccine in patients with severe leishmaniasis. Four patients with MCL and 3 with early DCL were treated with monthly intradermal injections of a vaccine containing promastigotes of Leishmania (Viannia braziliensis killed by pasteurization and viable Bacillus Calmette- Guerin. Clinical and immunological responses were evaluated. Integrity of protein constituents in extracts of pasteurized promastigotes was evaluated by gel electrophoresis. Complete remission of lesions occurred after 5-9 injections in patients with MCL or 7-10 injections in patients with early DCL. DCL patients developed positive skin reactions, average size 18.7 mm. All have been free of active lesions for at least 10 months. Adverse effects of the vaccine were limited to local reactivity to BCG at the injection sites and fever in 2 patients. Extracts of pasteurized and fresh promastigotes did not reveal differences in the integrity of protein components detectable by gel electrophoresis. Immunotherapy with this modified vaccine offers an effective, safe option for the treatment of patients who do not respond to immunotherapy with vaccine containing autoclaved parasites or to chemotherapy .

  17. Proceedings of the CSNI workshop on International Standard Problem 48 - Analysis of 1:4-scale prestressed concrete containment vessel model under severe accident conditions

    International Nuclear Information System (INIS)

    2005-01-01

    At the CSNI meeting in June 2002, the proposal for an International Standard Problem on containment integrity (ISP 48) based on the NRC/NUPEC/Sandia test was approved. Objectives were to extend the understanding of capacities of actual containment structures based on results of the recent PCCV Model test and other previous research. The ISP was sponsored by the USNRC, and results had been made available thanks to NUPEC and to the USNRC. Sandia National Laboratory was contracted to manage the technical aspects of the ISP. At the end of the ISP48, a workshop was organized in Lyon, France on April 6-7, 2005 hosted by Electricite de France. Its overall objective was to present results obtained by participants in the ISP 48 and to assess the current practices and the state of the art with respect to the calculation of concrete structures under severe accident conditions. Experience from other areas in civil engineering related to the modelling of complex structures was greatly beneficial to all. Information obtained as a result of this assessment were utilized to develop a consensus on these calculations and identify issues or 'gaps' in the present knowledge for the primary purpose of formulating and prioritizing research needs on this topic. The ISP48 exercise was published in the report referenced NEA/CSNI/R(2005)5 in 3 volumes. Volume 1 contains the synthesis of the exercise; Volumes 2 and 3 contain individual contributions of participating organizations. The CSNI Working Group on the Integrity and Ageing and in particular its sub-group on the behaviour of concrete structures has produced extensive material over the last few years. The complete list of references is given in this document. These proceedings gather the papers and presentations given by the participants at the Lyon workshop

  18. Insights into the control of the release of iodine, cesium, strontium and other fission products in the containment by severe accident management

    International Nuclear Information System (INIS)

    2000-03-01

    This document is intended to provide a management-level overview of the technical bases for accident management activities to attenuate releases of radioactive materials in the very unlikely event of a severe nuclear power reactor accident - activities known commonly as management of severe accident source terms. Such activities are natural complements to accident management activities directed at arresting or slowing accident progression. Abbreviated, qualitative discussions are presented in the document on the more important severe nuclear reactor accidents, the nature of radioactive material releases during accidents, natural processes that act to attenuate the amount of radioactive material that can escape a power plant, and the physical and chemical principles used in engineered systems to further attenuate radioactive releases during accidents. At the end of each section of the report, an annotated bibliography is provided. These bibliographies are intended to serve as introductions to the vast literature pertinent to all aspects of accident management including the management of radioactive source terms. Finally, it must be noted that much of the presentation has been made from the perspective of conventional pressurized water reactors and boiling water reactors. Many important details will be different for other types of reactors or for reactors with special features. Readers are asked to do the mental manipulations necessary to apply the ideas discussed here to the particular circumstances and features of their own reactors. The report is based on the following outline: - a brief discussion of fission product sources; fission product characteristics; chemical compounds; - transport and deposition of fission products; brief description of different deposition and agglomeration processes; - retention of fission products; re-evaporation, resuspension, etc.; - discussion of various possibilities to enhance the removal of fission products from the containment

  19. Vapour pressures, osmotic and activity coefficients for binary mixtures containing (1-ethylpyridinium ethylsulfate + several alcohols) at T = 323.15 K

    International Nuclear Information System (INIS)

    Calvar, Noelia; Gomez, Elena; Dominguez, Angeles; Macedo, Eugenia A.

    2010-01-01

    Osmotic coefficients of binary mixtures containing several primary and secondary alcohols (1-propanol, 2-propanol, 1-butanol, 2-butanol, and 1-pentanol) and the pyridinium-based ionic liquid 1-ethylpyridinium ethylsulfate were determined at T = 323.15 K using the vapour pressure osmometry technique. From the experimental results, vapour pressure and activity coefficients can be determined. For the correlation of osmotic coefficients, the extended Pitzer model modified by Archer, and the modified NRTL (MNRTL) model were used, obtaining deviations lower than 0.017 and 0.047, respectively. The mean molal activity coefficients and the excess Gibbs free energy for the binary mixtures studied were determined from the parameters obtained with the extended Pitzer model modified by Archer.

  20. CONTAIN code calculations of the effects on the source term of CsI to I/sub 2/ conversion due to severe hydrogen burns

    International Nuclear Information System (INIS)

    Valdez, G.D.; Williams, D.C.

    1986-01-01

    In experiments conducted at Sandia National Laboratories large amounts of elemental iodine were produced when CsI-Al 2 O 3 aerosol was exposed to hydrogen/air combustion. To evaluate some of the implications of the iodide conversion (observed to occur with up to 75% efficiency) for the severe accident source term, computational simulations of representative accident sequences were conducted with the CONTAIN code. The following conclusions can be drawn from this preliminary source term assessment: (1) If the containment sprays are inoperative during the accident, or failed by the hydrogen burn, the late-time source term is almost tripled when the iodide is converted to I 2 . (2) With the sprays active, the amount released without conversion of the CsI aerosol is 63% higher than for the case when conversion occurs. (3) For the case where CsI is converted to I 2 continued operation of the sprays reduces the release by a factor of 40, relative to the case in which the sprays fail at the time of the hydrogen burn. When there is no conversion, the reduction factor for continued spray operation is about a factor of 9, relative to the failed spray case

  1. 10 CFR 32.26 - Gas and aerosol detectors containing byproduct material: Requirements for license to manufacture...

    Science.gov (United States)

    2010-01-01

    ...: Requirements for license to manufacture, process, produce, or initially transfer. 32.26 Section 32.26 Energy NUCLEAR REGULATORY COMMISSION SPECIFIC DOMESTIC LICENSES TO MANUFACTURE OR TRANSFER CERTAIN ITEMS... byproduct material: Requirements for license to manufacture, process, produce, or initially transfer. An...

  2. Kiche: A simulation tool for kinetics of iodine chemistry in the containment of light water reactors under severe accident conditions (Contract research)

    International Nuclear Information System (INIS)

    Moriyama, Kiyofumi; Maruyama, Yu; Nakamura, Hideo

    2011-03-01

    An iodine chemistry simulation tool, Kiche, was developed for analyses of chemical kinetics relevant to iodine volatilization in the containment vessel of light water reactors (LWRs) during a severe accident. It consists of a Fortran code to solve chemical kinetics models, reaction databases written in plain text format, and peripheral tools to convert the reaction databases into Fortran codes to solve corresponding ordinary differential equation sets. Potential advantages of Kiche are the text format reaction database separated from the code that provides flexibility of the chemistry model, and, being a Fortran code which is relatively easily coupled with other Fortran codes such as severe accident analysis codes. This document describes the model, solution method, code structure, and examples of application of Kiche for simulation of experiments. The calculation results by the present model agreed well with the experimental data and it indicates the model properly includes the most important processes in the volatilization of iodine from irradiated iodide solutions with or without organic impurities. The appendixes give practical information for the usage of Kiche. (author)

  3. The effects of the combination of chlorhexidine/thymol- and fluoride-containing varnishes on the severity of root caries lesions in frail institutionalised elderly people.

    Science.gov (United States)

    Brailsford, S R; Fiske, J; Gilbert, S; Clark, D; Beighton, D

    2002-01-01

    To compare the clinical effects of a fluoride-containing varnish (Fluor-Protector) in combination with a chlorhexidine-containing varnish (Cervitec) on existing root caries lesions in a group of frail elderly subjects. A randomised double blind longitudinal study was utilised. Subjects (n = 102) were randomly allocated to a Test or Placebo group. All leathery and soft root caries lesions in all subjects were coated with Fluor-Protector while the lesions in the Test group were also coated with Cervitec and the lesions in the Placebo group were coated with a Placebo varnish. Treatments were repeated five times in a 12-month period. Clinical parameters associated with root caries, measurements of individual lesions and salivary levels of caries associated bacteria were made at intervals. The clinical severity of the lesions in the Test group did not change significantly during the 12-month study period. In the Placebo group the mean lesion width and lesion height and length of exposed root increased significantly and the lesions were significantly closer to the gingival margin. There were no significant changes in the salivary levels of caries-associated microorganisms after 12 months although, in both groups, there was initially a significant reduction in the salivary levels of mutans streptococci. The combination of Fluor-Protector and Cervitec is a useful, simple, quick and non-invasive method for the control and management of existing root caries lesions. The procedure could be performed by a dental hygienist and may be usefully applied in other high-risk groups including persons with Parkinson's disease, debilitating neuromuscular conditions and dry mouth from whatever cause. Copyright 2002 Elsevier Science Ltd.

  4. Compilation of Requirements for Safe Handling of Fluorine and Fluorine-Containing Products of Uranium Hexafluoride Conversion

    International Nuclear Information System (INIS)

    Ferrada, J.J.; Hightower, J.R.; Begovich, J.M.

    2000-01-01

    Public Law (PL) 105--204 requires the U.S. Department of Energy to develop a plan for inclusion in the fiscal year 2000 budget for conversion of the Department's stockpile of depleted uranium hexafluoride (DUF6) to a more stable form over an extended period. The conversion process into a more stable form will produce fluorine compounds (e.g., elemental fluorine or hydrofluoric acid) that need to be handled safely. This document compiles the requirements necessary to handle these materials within health and safety standards, which may apply in order to ensure protection of the environment and the safety and health of workers and the public

  5. Secondary Containers and Service Containers for Pesticides

    Science.gov (United States)

    Secondary containers and service containers are used by pesticide applicators in the process of applying a pesticide. EPA does not require secondary containers or service containers to be labeled or to meet particular construction standards. Learn more.

  6. Understanding of the operation behaviour of a Passive Autocatalytic Recombiner (PAR) for hydrogen mitigation in realistic containment conditions during a severe Light Water nuclear Reactor (LWR) accident

    International Nuclear Information System (INIS)

    Payot, Frédéric; Reinecke, Ernst-Arndt; Morfin, Franck; Sabroux, Jean-Christophe; Meynet, Nicolas; Bentaib, Ahmed; March, Philippe; Zeyen, Roland

    2012-01-01

    Highlights: ► Recombineur operation in the presence of fission products (severe accident conditions). ► Operation of catalysts in the integral and small-scale tests. ► The catalyst performance was observed by measuring the coupon temperature increase. ► The experimental observations were corroborated by numerical calculations (SPARK code). - Abstract: In the context of hydrogen risk mitigation in nuclear power plants (NPPs), experimental studies of a possible poisoning of Passive Autocatalytic Recombiners (PARs) by fission products (FPs) and aerosols released during a core meltdown accident were mainly conducted in the past with non-radioactive fission product surrogates (e.g., in the H2PAR facility at Cadarache, France). The decision was taken in 1997 to complete these studies by a test in the Phébus facility, a research nuclear reactor also at Cadarache: it was a rare opportunity to expose catalyst samples to an atmosphere as representative as possible of a core meltdown accident, containing gaseous fission products and aerosols released during the degradation of an actual irradiated nuclear fuel bundle. Before testing in Phébus during the FPT3 experiment, reference and qualification tests were performed in the H2PAR facility using the same samples — the so-called “coupons” — and coupons holder to check that the apparatus was functional and correctly designed for avoiding to tamper with the thermal-hydraulics and chemical conditions in the Phébus containment. The correct operation of catalysts was checked by measuring the surface temperature increase of the coupons due to the exothermic reaction between hydrogen and oxygen. After the Phébus FPT3 test (November 2004), REKO-1 tests were initiated at Jülich, Germany, to confirm the discrepancy in coupons temperature observed in Phébus FPT3 and H2PAR PHEB-03 tests, and to study the operation behaviour of PARs. Besides, before REKO-1 tests, a first interpretation of H2PAR and Phébus experiments

  7. Greater-than-Class C low-level radioactive waste shipping package/container identification and requirements study. National Low-Level Waste Management Program

    Energy Technology Data Exchange (ETDEWEB)

    Tyacke, M.

    1993-08-01

    This report identifies a variety of shipping packages (also referred to as casks) and waste containers currently available or being developed that could be used for greater-than-Class C (GTCC) low-level waste (LLW). Since GTCC LLW varies greatly in size, shape, and activity levels, the casks and waste containers that could be used range in size from small, to accommodate a single sealed radiation source, to very large-capacity casks/canisters used to transport or dry-store highly radioactive spent fuel. In some cases, the waste containers may serve directly as shipping packages, while in other cases, the containers would need to be placed in a transport cask. For the purpose of this report, it is assumed that the generator is responsible for transporting the waste to a Department of Energy (DOE) storage, treatment, or disposal facility. Unless DOE establishes specific acceptance criteria, the receiving facility would need the capability to accept any of the casks and waste containers identified in this report. In identifying potential casks and waste containers, no consideration was given to their adequacy relative to handling, storage, treatment, and disposal. Those considerations must be addressed separately as the capabilities of the receiving facility and the handling requirements and operations are better understood.

  8. Phase equilibrium properties of binary aqueous solutions containing ethanediamine, 1,2-diaminopropane, 1,3-diaminopropane, or 1,4-diaminobutane at several temperatures

    Energy Technology Data Exchange (ETDEWEB)

    Ahmed, Nouria Chiali-Baba [LATA2M, Laboratoire de Thermodynamique Appliquee et Modelisation Moleculaire, University AbouBekr Belkaid of Tlemcen, P.O. Box 119, Tlemcen 13000 (Algeria); Negadi, Latifa, E-mail: l_negadi@mail.univ-tlemcen.d [LATA2M, Laboratoire de Thermodynamique Appliquee et Modelisation Moleculaire, University AbouBekr Belkaid of Tlemcen, P.O. Box 119, Tlemcen 13000 (Algeria); Mokbel, Ilham; Jose, Jacques [LSA, Laboratoire des Sciences Analytiques, CNRS-UMR 5180, Universite Claude Bernard - Lyon I. 43, Bd du 11 Novembre 1918, Villeurbanne Cedex 69622 (France)

    2011-05-15

    Research highlights: Vapour pressures of ethanediamine (EDA), 1,2-diaminopropane, 1,3-diaminopropane (1,3-DAP), or 1,4-diaminobutane (1,4-DAB) aqueous solutions are reported between (293 and 363) K. The two first mixtures show negative azeotropic behaviour. The aqueous solutions of EDA, 1,2-DAP, or 1,3-DAP exhibit negative G{sup E} whereas the one containing 1,4-DAB shows either negative G{sup E} or sinusoidal shape for G{sup E}. - Abstract: The vapour pressures of {l_brace}ethanediamine (EDA) + water{r_brace}, {l_brace}1,2-diaminopropane (1,2-DAP) + water{r_brace}, {l_brace}1,3-diaminopropane (1,3-DAP) + water{r_brace} or {l_brace}1,4-diaminobutane (1,4-DAB) + water{r_brace} binary mixtures, and of pure EDA, 1,2-DAP, 1,3-DAP, 1,4-DAB, and water components were measured by means of two static devices at temperatures between (293 and 363) K. The data were correlated with the Antoine equation. From these data, the excess Gibbs function (G{sup E}) was calculated for several constant temperatures and fitted to a fourth-order Redlich-Kister equation using the Barker's method. The {l_brace}ethanediamine (EDA) + water{r_brace}, and {l_brace}1,2-diaminopropane (1,2-DAP) + water{r_brace} binary systems show negative azeotropic behaviour. The aqueous solutions of EDA, 1,2-DAP, or 1,3-DAP exhibit negative deviations in G{sup E} for all investigated temperatures over the whole composition range whereas the (1,4-DAB + water) binary mixture shows negative G{sup E} for temperatures (293.15 < T/K < 353.15) and a sinusoidal shape for G{sup E} at T = 363.15 K.

  9. Final environmental statement concerning rule making. Exemption from licensing requirements for spark-gap irradiators that contain cobalt-60. Docket No. PRM 30-54

    International Nuclear Information System (INIS)

    1977-12-01

    The potential environmental impacts and adverse environmental effects from distribution, use only in commercial-sized oil burners, and disposal of 6000 spark-gap irradiators per year that contain 60 Co are summarized. On the basis of the analysis and evaluation set forth in this statement and after weighing the environmental, economic, technical, and other benefits against environmental costs and after considering available alternatives, it is concluded that the action called for under the National Environmental Policy Act of 1969 (NEPA) and 10 CFR Part 51 is the issuance of an exemption from licensing requirements for spark-gap irradiators that contain 60 Co, subject to the following conditions for the protection of the environment: persons who apply 60 Co to, or persons who incorporate 60 Co into, spark-gap irradiators or persons who import for sale or distribution spark-gap irradiators containing 60 Co are not exempt from the requirements for a license; each spark-gap irradiator shall contain no more than 1 μCi of 60 Co; and the 60 Co shall be applied to the spark-gap irradiators for use in electrically ignited fuel-oil burners having a firing rate of at least 3 gal/h

  10. Calculation of mean outcrossing rates of non-Gaussian processes with stochastic input parameters - Reliability of containers stowed on ships in severe sea

    DEFF Research Database (Denmark)

    Nielsen, Ulrik Dam

    2010-01-01

    values is expected to occur, and the final result, the mean outcrossing rate, is obtained by summation. The derived procedure is illustrated by an example considering the forces in containers stowed on ships and, in particular, results are presented for the so-called racking failure in the containers...

  11. Maintaining Aura's Orbit Requirements While Performing Orbit Maintenance Maneuvers Containing an Orbit Normal Delta-V Component

    Science.gov (United States)

    Johnson, Megan R.; Petersen, Jeremy D.

    2014-01-01

    The Earth Observing System (EOS) Afternoon Constellation consists of five member missions (GCOM-W1, Aqua, CALIPSO, CloudSat, and Aura), each of which maintain a frozen, sun-synchronous orbit with a 16-day repeating ground track that follows the Worldwide Reference System-2 (WRS-2). Under nominal science operations for Aura, the propulsion system is oriented such that the resultant thrust vector is aligned 13.493 degrees away from the velocity vector along the yaw axis. When performing orbit maintenance maneuvers, the spacecraft performs a yaw slew to align the thrust vector in the appropriate direction. A new Drag Make Up (DMU) maneuver operations scheme has been implemented for Aura alleviating the need for the 13.493 degree yaw slew. The focus of this investigation is to assess the impact that no-slew DMU maneuver operations will have on Aura's Mean Local Time (MLT) which drives the required along track separation between Aura and the constellation members, as well as Aura's frozen orbit properties, eccentricity and argument of perigee. Seven maneuver strategies were analyzed to determine the best operational approach. A mirror pole strategy, with maneuvers alternating at the North and South poles, was implemented operationally to minimize impact to the MLT. Additional analysis determined that the mirror pole strategy could be further modified to include frozen orbit maneuvers and thus maintain both MLT and the frozen orbit properties under noslew operations.

  12. University course timetabling and the requirements: Survey in several universities in the east-coast of Malaysia

    Science.gov (United States)

    Aziz, Nurul Liyana Abdul; Aizam, Nur Aidya Hanum

    2017-08-01

    Course timetabling problem receives the highlight at the beginning of every semester. The problem is mainly on assigning courses to timeslot, rooms and lecturers which involving a set of rules and policies constraints. Generally, researchers present different features to signify their own universities' timetable according to the structure and behavior of their institution. However, the gap between theory and real-world applications that can be seen in the resulted timetable is the lacking of acknowledging human preferences. As to overcome this, it is very important to consider all the demands and preferences from timetabling community. This research therefore tries to accommodate the problem by investigating through surveys to several universities in the east coast of Malaysia the demands and preferences of individuals involved directly. Results from the questionnaires will be analyzed by using SPSS and all current issues regarding the demands will be included into our existing general university course timetabling mathematical model. The new university course timetabling mathematical model could best represent universities and be useful, especially in universities in Malaysia.

  13. French regulatory requirements for the occupational radiation protection in severe accident situations and post-accident recovery

    International Nuclear Information System (INIS)

    Couasnon, Olivier

    2014-01-01

    -accident), intervention personnel receive radiation protection granted to exposed workers. ASN will have to take into account two major sources of implementation of the occupational radiation protection during an emergency situation: the transposition of Council Directive 2013/59/EURATOM of 5 December 2013 and the requirements following the complementary safety assessments of the nuclear power plants in the light of the accident that occurred on the nuclear power plant at Fukushima Daiichi. Indeed, member States shall bring into force the laws, regulations and administrative provisions necessary to comply with the Directive. For example, in the French regulation, the end of the emergency situation and the transition from emergency phase to the recovery phase are not mentioned and will have to be integrated in the French legal framework. Concerning the complementary safety assessments, they require a 'hard core' of material and organizational measures designed to ensure control of basic safety functions in extreme situations (comprising operational dosimetry resources for workers) and in addition that the operator (EDF) gradually deploy its proposed national 'Nuclear rapid response force (FARN)' comprising specialist crews and equipment able to take over from the personnel on a site affected by an accident. (author)

  14. A pilot outreach physiotherapy and dietetic quality improvement initiative reduces IV antibiotic requirements in children with moderate-severe cystic fibrosis.

    Science.gov (United States)

    Ledger, Sean J; Owen, Elizabeth; Prasad, S Ammani; Goldman, Allan; Willams, Jane; Aurora, Paul

    2013-12-01

    At our hospital the current model of care for children with moderate-severe CF is focused on intensive inpatient intervention, regular outpatient clinic review and specialist outreach care as required. An alternative model providing more regular physiotherapy and dietetic outreach support, in addition to these specialist services, may be more effective. 16 children (4 male; 12 female; mean age 10.9±2.93; range 4-15 years) who required >40days of IV antibiotics in the 12-months pre-intervention were enrolled. Physiotherapy included weekly-supervised exercise sessions, alongside regular review of home physiotherapy regimens. Dietetic management included 1-2 monthly monitoring of growth, appetite, intake and absorption, and nutrition education sessions. There was a 23% reduction in inpatient IV antibiotic requirement and 20% reduction in home IV antibiotic requirement during the intervention year. Cost-benefit analyses showed savings of £113,570. VO(2Peak) increased by 4.9 ml·kg·min(-1) (95%CI 1.01 to 8.71; p=0.02), and 10 m-MSWT distance and increment achieved increased by 229 m (95%CI 109 to 350; pchildren with moderate-severe CF. A fully powered clinical trial is now warranted. Copyright © 2013 European Cystic Fibrosis Society. Published by Elsevier B.V. All rights reserved.

  15. Medicago truncatula DNF2 is a PI-PLC-XD-containing protein required for bacteroid persistence and prevention of nodule early senescence and defense-like reactions.

    Science.gov (United States)

    Bourcy, Marie; Brocard, Lysiane; Pislariu, Catalina I; Cosson, Viviane; Mergaert, Peter; Tadege, Millon; Mysore, Kirankumar S; Udvardi, Michael K; Gourion, Benjamin; Ratet, Pascal

    2013-03-01

    Medicago truncatula and Sinorhizobium meliloti form a symbiotic association resulting in the formation of nitrogen-fixing nodules. Nodule cells contain large numbers of bacteroids which are differentiated, nitrogen-fixing forms of the symbiotic bacteria. In the nodules, symbiotic plant cells home and maintain hundreds of viable bacteria. In order to better understand the molecular mechanism sustaining the phenomenon, we searched for new plant genes required for effective symbiosis. We used a combination of forward and reverse genetics approaches to identify a gene required for nitrogen fixation, and we used cell and molecular biology to characterize the mutant phenotype and to gain an insight into gene function. The symbiotic gene DNF2 encodes a putative phosphatidylinositol phospholipase C-like protein. Nodules formed by the mutant contain a zone of infected cells reduced to a few cell layers. In this zone, bacteria do not differentiate properly into bacteroids. Furthermore, mutant nodules senesce rapidly and exhibit defense-like reactions. This atypical phenotype amongst Fix(-) mutants unravels dnf2 as a new actor of bacteroid persistence inside symbiotic plant cells. © 2012 CNRS. New Phytologist © 2012 New Phytologist Trust.

  16. Atg6/UVRAG/Vps34-Containing Lipid Kinase Complex Is Required for Receptor Downregulation through Endolysosomal Degradation and Epithelial Polarity during Drosophila Wing Development

    Directory of Open Access Journals (Sweden)

    Péter Lőrincz

    2014-01-01

    Full Text Available Atg6 (Beclin 1 in mammals is a core component of the Vps34 PI3K (III complex, which promotes multiple vesicle trafficking pathways. Atg6 and Vps34 form two distinct PI3K (III complexes in yeast and mammalian cells, either with Atg14 or with UVRAG. The functions of these two complexes are not entirely clear, as both Atg14 and UVRAG have been suggested to regulate both endocytosis and autophagy. In this study, we performed a microscopic analysis of UVRAG, Atg14, or Atg6 loss-of-function cells in the developing Drosophila wing. Both autophagy and endocytosis are seriously impaired and defective endolysosomes accumulate upon loss of Atg6. We show that Atg6 is required for the downregulation of Notch and Wingless signaling pathways; thus it is essential for normal wing development. Moreover, the loss of Atg6 impairs cell polarity. Atg14 depletion results in autophagy defects with no effect on endocytosis or cell polarity, while the silencing of UVRAG phenocopies all but the autophagy defect of Atg6 depleted cells. Thus, our results indicate that the UVRAG-containing PI3K (III complex is required for receptor downregulation through endolysosomal degradation and for the establishment of proper cell polarity in the developing wing, while the Atg14-containing complex is involved in autophagosome formation.

  17. Containment Evaluation under Severe Accidents (CESA): synthesis of the predictive calculations and analysis of the first experimental results obtained on the Civaux mock-up

    International Nuclear Information System (INIS)

    Granger, L.; Rieg, C.Y.; Touret, J.P.; Fleury, F.; Nahas, G.; Danisch, R.; Brusa, L.; Millard, A.; Laborderie, C.; Ulm, F.; Contri, P.; Schimmelpfennig, K.; Barre, F.; Firnhaber, M.; Gauvain, J; Coulon, N.; Dutton, L.M.C.; Tuson, A.

    2001-01-01

    In 1996, EDF decided to build a containment model at the scale 1:3, the MAEVA mock-up, in order to check and study the behaviour of a pre-stressed concrete containment vessel without a liner in terms of mechanical strength and leaktightness, for loadings corresponding to its design and beyond design conditions. In parallel with the construction and testing of the mock-up, a cost-shared R and D action supported by the European Union, the CESA project, is dealing with quantification of leak rates through concrete cracks and porosity, predictive calculations of the behaviour of the mock-up and analysis of the experimental results. In this paper, we propose a synthesis of the main theoretical and experimental results, obtained after 2.5 years. It should however be noted that, due to some unexpected delays in the experimental programme, quite natural with such a huge and unique experimental set-up, only the design-basis accident sequences, already performed, have been reported in this paper. The first results are nevertheless very interesting, both from a scientific and nuclear utility point of view

  18. Safety and Efficacy of an Artificial Tear Containing 0.3% Hyaluronic Acid in the Management of Moderate-to-Severe Dry Eye Disease.

    Science.gov (United States)

    López-de la Rosa, Alberto; Pinto-Fraga, José; Blázquez Arauzo, Francisco; Urbano Rodríguez, Rubén; González-García, María J

    2017-11-01

    To evaluate the safety and efficacy of a new 0.3% hyaluronic acid artificial tear compared with 0.9% saline solution (0.9% NaCl) in moderate-to-severe dry eye patients after 1 month's use. A total of 16 patients with moderate-to-severe dry eye were included in this crossover study. After a 1-week washout period, patients used the experimental (Visaid 0.3%) or control solution (0.9% NaCl), selected randomly, applying three to eight drops daily for a month. After another washout period, patients used the other solution in the same way. Percentage of change (ΔY) was calculated and analyzed for (1) safety variables: visual acuity, intraocular pressure, and ophthalmoscopy evaluation; (2) efficacy variable: Ocular Surface Disease Index (OSDI) questionnaire; and (3) secondary variables: biomicroscopy findings, fluorescein corneal staining, lissamine green conjunctival staining, tear breakup time (TBUT), contrast sensitivity, Schirmer test, and subject satisfaction. There were no significant differences in the safety parameters for either solution. After using Visaid 0.3%, patients showed significant improvements in OSDI score (ΔY: -9.66%±10.90), tarsal hyperemia (ΔY: -16.67%±27.89), corneal staining extension (ΔY: -34.90%±42.41), TBUT (ΔY: 13.98%±26.19), and subjective satisfaction (ΔY: 38.06%±47.06). When using 0.9% NaCl, Schirmer test results were significantly worse (ΔY: -11.47%±19.27). A significant difference between the 2 solutions was found in TBUT (ΔY: 13.98%±26.19 vs. 10.15%±42.34, respectively; P=0.0214). Visaid 0.3% is a safe product with some benefits over 0.9% NaCl in reducing ocular symptoms and improving some ocular signs in patients with moderate-to-severe dry eye.

  19. A DNA fragment from Xq21 replaces a deleted region containing the entire FVIII gene in a severe hemophilia A patient

    Energy Technology Data Exchange (ETDEWEB)

    Murru, S.; Casula, L.; Moi, P. [Insituto di Clinica e Biologia dell` Eta Evolutiva, Cagliari (Italy)] [and others

    1994-09-15

    In this paper the authors report the molecular characterization of a large deletion that removes the entire Factor VIII gene in a severe hemophilia A patient. Accurate DNA analysis of the breakpoint region revealed that a large DNA fragment replaced the 300-kb one, which was removed by the deletion. Pulsed-field gel electrophoresis analysis revealed that the size of the inserted fragment is about 550 kb. In situ hybridization demonstrated that part of the inserted region normally maps to Xq21 and to the tip of the short arm of the Y chromosome (Yp). In this patient this locus is present both in Xq21 and in Xq28, in addition to the Yp, being thus duplicated in the X chromosome. Sequence analysis of the 3` breakpoint suggested that an illegitimate recombination is probably the cause of this complex rearrangement. 52 refs., 7 figs.

  20. Thermodynamic Modeling of Several Aqueous Alkanol Solutions Containing Amino Acids with the Perturbed-Chain Statistical Associated Fluid Theory Equation of State

    DEFF Research Database (Denmark)

    Ferreira, Luisa; Breil, Martin Peter; Pinho, S. P.

    2009-01-01

    parameters for the amino acids were fitted to the densities, activity and osmotic coefficients, vapor pressures, and water activity of their aqueous solutions. The solubilities of amino acids in pure and mixed solvent systems were calculated on the basis of the phase equilibrium conditions for a pure solid...... and a fluid phase. The hypothetical melting properties of each amino acid were fitted, to accurately correlate the solubilities in pure water. Only one temperature independent binary parameter is required for each amino acid/solvent pair. The model can accurately describe the solubility of the amino acids...... in water, but the correlation for the solubility in pure alcohols was not so satisfactory. The solubility in mixed solvents (ternary systems) was predicted on the basis of the modeling of the solubility in pure solvents, without any additional fitting of the parameters, and the results achieved were...

  1. Lipidation of BmAtg8 is required for autophagic degradation of p62 bodies containing ubiquitinated proteins in the silkworm, Bombyx mori.

    Science.gov (United States)

    Ji, Ming-Ming; Lee, Jae Man; Mon, Hiroaki; Iiyama, Kazuhiro; Tatsuke, Tsuneyuki; Morokuma, Daisuke; Hino, Masato; Yamashita, Mami; Hirata, Kazuma; Kusakabe, Takahiro

    2017-10-01

    p62/Sequestosome-1 (p62/SQSTM1, hereafter referred to as p62) is a major adaptor that allows ubiquitinated proteins to be degraded by autophagy, and Atg8 homologs are required for p62-mediated autophagic degradation, but their relationship is still not understood in Lepidopteran insects. Here it is clearly demonstrated that the silkworm homolog of mammalian p62, Bombyx mori p62 (Bmp62), forms p62 bodies depending on its Phox and Bem1p (PB1) and ubiquitin-associated (UBA) domains. These two domains are associated with Bmp62 binding to ubiquitinated proteins to form the p62 bodies, and the UBA domain is essential for the binding, but Bmp62 still self-associates without the PB1 or UBA domain. The p62 bodies in Bombyx cells are enclosed by BmAtg9-containing membranes and degraded via autophagy. It is revealed that the interaction between the Bmp62 AIM motif and BmAtg8 is critical for the autophagic degradation of the p62 bodies. Intriguingly, we further demonstrate that lipidation of BmAtg8 is required for the Bmp62-mediated complete degradation of p62 bodies by autophagy. Our results should be useful in future studies of the autophagic mechanism in Lepidopteran insects. Copyright © 2017 Elsevier Ltd. All rights reserved.

  2. Requirements to be taken into account when designing safety-related mechanical components conveying or containing pressurized fluid and classified as level 2 or 3

    International Nuclear Information System (INIS)

    1984-12-01

    RFS or Regles Fondamentales de Surete (Basic Safety Rules) applicable to certain types of nuclear facilities lay down requirements with which compliance, for the type of facilities and within the scope of application covered by the RFS, is considered to be equivalent to compliance with technical French regulatory practice. The object of the RFS is to take advantage of standardization in the field of safety, while allowing for technical progress in that field. They are designed to enable the operating utility and contractors to know the rules pertaining to various subjects which are considered to be acceptable by the Service Central de Surete des Installations Nucleaires, or the SCSIN (Central Department for the Safety of Nuclear Facilities). These RFS should make safety analysis easier and lead to better understanding between experts and individuals concerned with the problems of nuclear safety. The SCSIN reserves the right to modify, when considered necessary, any RFS and specify, if need be, the terms under which a modification is deemed retroactive. The purpose of this RFS is to specify the requirements to be taken into account when designing mechanical components conveying or containing pressurized fluid and which are in safety class 2 or 3

  3. Study of the performance of gels of molybdenum containing several cations for the preparation of 99Mo and 99mTc

    International Nuclear Information System (INIS)

    Moraes, Vanessa

    2005-01-01

    99m Tc is the most employed radioisotope in Nuclear Medicine, due to its nuclear characteristics: short half-life (6.04 h); emission of low energy gamma ray (140 keV); no emission of β - ; generated by the radioactive decay of 99 Mo (radioisotope generator system). 99 Mo can be produced in cyclotron or nuclear reactor by the irradiation of 235 U (n, f) 99 Mo or by the 98 Mo (n, γ) 90 Mo reaction. Four different kinds of generators of 99m Tc can be employed, based on the separation techniques: column chromatographic using alumina, with fission 99 Mo; solvent extraction using methylethylketone; sublimation of technetium heptoxide; gel type chromatographic generator, that contains molybdenum. IPEN, aiming the nationalization of the 99m Tc generators production, developed a gel type generator that uses zirconium molybdate. Three types of gels are studied in the work: molybdenum gel with titanium, molybdenum gel with cerium and molybdenum gel with hafnium, that were compared with the molybdenum gel with zirconium. The variables studied in the gel preparation are: mass relation between Mo and the cation, NaOH concentration, temperature and final pH of the product. After the preparation, the gels are analysed in relation to the amount of Mo and the cation, structure and gel particle size. The gel is irradiated and later a generator system is prepared, and the elutions are analysed in order to measure the 99m Tc elution efficiency. The results showed that the molybdenum gel with titanium had the best performance in all analysis. (author)

  4. Review of current severe accident management approaches in Europe and identification of related modelling requirements for the computer code ASTEC V2.1

    International Nuclear Information System (INIS)

    Hermsmeyer, S.

    2015-01-01

    The severe accident at the Fukushima-Daiichi nuclear power plant (NPP) has led to a worldwide review of nuclear safety approaches and is bringing a refocussing of R and D in the field. To support these efforts several new Euratom FP7 projects have been launched. The CESAM project focuses on the improvement of the ASTEC computer code. ASTEC is jointly developed by IRSN and GRS and is considered as the European reference code for Severe Accident Analyses since it capitalizes knowledge from the extensive Euro-pean R and D in the field. The project aims at the code's enhancement and extension for use in Severe Accident Management (SAM) analysis of the NPPs of Generation II-III presently under operation or foreseen in the near future in Europe, spent fuel pools included. The work reported here is concerned with the importance, for the further development of the code, of SAM strategies to be simulated. To this end, SAM strategies applied in the EU have been compiled. This compilation is mainly based on the public information made available in the frame of the EU ''stress tests'' for NPPs and has been complemented by information pro-vided by the different CESAM partners. The context of SAM is explained and the strategies are presented. The modelling capabilities for the simulation of these strategies in the current production version 2.0 of ASTEC are discussed. Furthermore, the requirements for the next version of ASTEC V2.1 that is supported in the CESAM project are highlighted. They are a necessary complement to the list of code improvements that is drawn from consolidating new fields of application, like SFP and BWR model enhancements, and from new experimental results on severe accident phenomena.

  5. Review of current severe accident management approaches in Europe and identification of related modelling requirements for the computer code ASTEC V2.1

    Energy Technology Data Exchange (ETDEWEB)

    Hermsmeyer, S. [European Commission JRC, Petten (Netherlands). Inst. for Energy and Transport; Herranz, L.E.; Iglesias, R. [CIEMAT, Madrid (Spain); and others

    2015-07-15

    The severe accident at the Fukushima-Daiichi nuclear power plant (NPP) has led to a worldwide review of nuclear safety approaches and is bringing a refocussing of R and D in the field. To support these efforts several new Euratom FP7 projects have been launched. The CESAM project focuses on the improvement of the ASTEC computer code. ASTEC is jointly developed by IRSN and GRS and is considered as the European reference code for Severe Accident Analyses since it capitalizes knowledge from the extensive Euro-pean R and D in the field. The project aims at the code's enhancement and extension for use in Severe Accident Management (SAM) analysis of the NPPs of Generation II-III presently under operation or foreseen in the near future in Europe, spent fuel pools included. The work reported here is concerned with the importance, for the further development of the code, of SAM strategies to be simulated. To this end, SAM strategies applied in the EU have been compiled. This compilation is mainly based on the public information made available in the frame of the EU ''stress tests'' for NPPs and has been complemented by information pro-vided by the different CESAM partners. The context of SAM is explained and the strategies are presented. The modelling capabilities for the simulation of these strategies in the current production version 2.0 of ASTEC are discussed. Furthermore, the requirements for the next version of ASTEC V2.1 that is supported in the CESAM project are highlighted. They are a necessary complement to the list of code improvements that is drawn from consolidating new fields of application, like SFP and BWR model enhancements, and from new experimental results on severe accident phenomena.

  6. Relationship between seasonal weather changes, risk of dehydration, and incidence of severe bradyarrhythmias requiring urgent temporary transvenous cardiac pacing in an elderly population

    Science.gov (United States)

    Palmisano, Pietro; Accogli, Michele; Zaccaria, Maria; Vergari, Alessandra; De Luca De Masi, Gabriele; Negro, Luca; De Blasi, Sergio

    2014-09-01

    There is little information on any seasonal variations or meteorological factors associated with symptomatic bradyarrhythmias requiring cardiac pacing. The aim of this single-center study was to investigate the seasonal distribution of the incidence of severe, life-threatening bradyarrhythmias requiring urgent temporary transvenous cardiac pacing in an elderly population. Consecutive patients who underwent urgent temporary transvenous cardiac pacing between 2007 and 2012 were enrolled. The baseline characteristics of the patients and some meteorological parameters, including the calculation the daily heat index (HI), were recorded. During the study period, 79 consecutive patients (mean age 82 ± 8 years, 41 % male) underwent urgent temporary transvenous cardiac pacing, mainly for third-degree atrioventricular block (79 %). The incidence of bradyarrhythmias was significantly higher in summer than in the other seasons ( P 90 °F for >3 h per day for at least 10 days ( P renal function impairment and hyperkalemia (all P < 0.05). This study showed an increased incidence of severe bradyarrhythmias in an elderly population during the hottest months of the year. In these months, in subjects characterized by increased susceptibility to dehydration, the risk of developing bradyarrhythmias was increased significantly.

  7. Severe Hyperammonemic Encephalopathy Requiring Dialysis Aggravated by Prolonged Fasting and Intermittent High Fat Load in a Ramadan Fasting Month in a Patient with CPTII Homozygous Mutation.

    Science.gov (United States)

    Phowthongkum, P; Ittiwut, C; Shotelersuk, V

    2017-11-21

    Carnitine palmitoyltransferase II (CPTII) deficiency is a mitochondrial fatty acid oxidation disorder that can present antenatally as congenital brain malformations, or postnatally with lethal neonatal, severe infantile, or the most common adult myopathic forms. No case of severe hyperammonemia without liver dysfunction has been reported. We described a 23-year-old man who presented to the emergency department with seizures and was found to have markedly elevation of serum ammonia. Continuous renal replacement therapy was initiated with successfully decreased ammonia to a safety level. He had a prolonged history of epilepsies and encephalopathic attacks that was associated with high ammonia level. Molecular diagnosis revealed a homozygous mutation in CPTII. The plasma acylcarnitine profile was consistent with the diagnosis. Failure to produce acetyl-CoA, the precursor of urea cycle from fatty acid in prolonged fasting state in Ramadan month, worsening mitochondrial functions from circulating long chain fatty acid and valproate toxicities were believed to contribute to this critical metabolic decompensation. Fatty acid oxidation disorders should be considered in the differential diagnosis of hyperammonemia even without liver dysfunction. To our knowledge, this is the first case of CPTII deficiency presented with severe hyperammonemic encephalopathy required dialysis after prolonged religious related fasting.

  8. [Efficacy of high versus low plasma: red blood cell ratio resuscitation in patients with severe trauma requiring massive blood transfusion: a meta-analysis].

    Science.gov (United States)

    Yu, Fang; Zhong, Tao; Wu, Gang

    2017-01-20

    To evaluate the efficacy of high (≥1:2) and low (ratio resuscitation in patients with severe trauma requiring massive blood transfusion. The databases including the Cochrane Library, Pubmed, Web of Science, and EMBASE were systemically searched for relevant studies published between January, 2009 and April, 2016. The selection of studies, assessment of methodological quality and data extraction were performed by two researchers independently according to the inclusion and exclusion criteria. The main endpoint was 24-h mortality, 30-day mortality and 24-h survival rate. Five observational studies reporting outcomes of 1024 patients were included in this meta-analysis. Four studies documented civilian cases and one study had a military setting. No significant differences were found in the Injury Severity Score (ISS) between patient groups receiving high and low plasma: RBC ratio resuscitation. Compared with the low-ratio group, the patients with high-ratio resuscitation showed a significant reduction in the 24-h mortality rate (OR=0.35, 95%CI [0.25, 0.48], Pratio resuscitation within the initial 24 h following the trauma (HR=2.34, 95%CI [1.46, 3.73], P=0.00001). Raising the plasma: RBC ratio to 0.5 or higher may decrease the mortality rate of the patients with severe trauma who need massive blood transfusion.

  9. BWR steel containment corrosion

    Energy Technology Data Exchange (ETDEWEB)

    Tan, C.P.; Bagchi, G.

    1996-04-01

    The report describes regulatory actions taken after corrosion was discovered in the drywell at the Oyster Creek Plant and in the torus at the Nine Mile Point 1 Plant. The report describes the causes of corrosion, requirements for monitoring corrosion, and measures to mitigate the corrosive environment for the two plants. The report describes the issuances of generic letters and information notices either to collect information to determine whether the problem is generic or to alert the licensees of similar plants about the existence of such a problem. Implementation of measures to enhance the containment performance under severe accident conditions is discussed. A study by Brookhaven National Laboratory (BNL) of the performance of a degraded containment under severe accident conditions is summarized. The details of the BNL study are in the appendix to the report.

  10. Regulation of abiotic stress signalling by Arabidopsis C-terminal domain phosphatase-like 1 requires interaction with a k-homology domain-containing protein.

    Directory of Open Access Journals (Sweden)

    In Sil Jeong

    Full Text Available Arabidopsis thaliana CARBOXYL-TERMINAL DOMAIN (CTD PHOSPHATASE-LIKE 1 (CPL1 regulates plant transcriptional responses to diverse stress signals. Unlike typical CTD phosphatases, CPL1 contains two double-stranded (ds RNA binding motifs (dsRBMs at its C-terminus. Some dsRBMs can bind to dsRNA and/or other proteins, but the function of the CPL1 dsRBMs has remained obscure. Here, we report identification of REGULATOR OF CBF GENE EXPRESSION 3 (RCF3 as a CPL1-interacting protein. RCF3 co-purified with tandem-affinity-tagged CPL1 from cultured Arabidopsis cells and contains multiple K-homology (KH domains, which were predicted to be important for binding to single-stranded DNA/RNA. Yeast two-hybrid, luciferase complementation imaging, and bimolecular fluorescence complementation analyses established that CPL1 and RCF3 strongly associate in vivo, an interaction mediated by the dsRBM1 of CPL1 and the KH3/KH4 domains of RCF3. Mapping of functional regions of CPL1 indicated that CPL1 in vivo function requires the dsRBM1, catalytic activity, and nuclear targeting of CPL1. Gene expression profiles of rcf3 and cpl1 mutants were similar during iron deficiency, but were distinct during the cold response. These results suggest that tethering CPL1 to RCF3 via dsRBM1 is part of the mechanism that confers specificity to CPL1-mediated transcriptional regulation.

  11. Closure requirements

    International Nuclear Information System (INIS)

    Hutchinson, I.P.G.; Ellison, R.D.

    1992-01-01

    Closure of a waste management unit can be either permanent or temporary. Permanent closure may be due to: economic factors which make it uneconomical to mine the remaining minerals; depletion of mineral resources; physical site constraints that preclude further mining and beneficiation; environmental, regulatory or other requirements that make it uneconomical to continue to develop the resources. Temporary closure can occur for a period of several months to several years, and may be caused by factors such as: periods of high rainfall or snowfall which prevent mining and waste disposal; economic circumstances which temporarily make it uneconomical to mine the target mineral; labor problems requiring a cessation of operations for a period of time; construction activities that are required to upgrade project components such as the process facilities and waste management units; and mine or process plant failures that require extensive repairs. Permanent closure of a mine waste management unit involves the provision of durable surface containment features to protect the waters of the State in the long-term. Temporary closure may involve activities that range from ongoing maintenance of the existing facilities to the installation of several permanent closure features in order to reduce ongoing maintenance. This paper deals with the permanent closure features

  12. Performance of the primary containment of a BWR during a severe accident whit the code RELAP/SCDAPSIM; Comportamiento del contenedor primario de un reactor BWR durante un accidente severo con el codigo RELAP/SCDAPSIM

    Energy Technology Data Exchange (ETDEWEB)

    Castillo G, F.

    2015-07-01

    In this thesis work, it was developed a model of the vacuum breaker valves and down comers for a BWR Mark II primary containment for the code RELAP/SCDAPSIM Mod. 3.4. This code was used to simulate a Station Blackout (Sbo) that evolves to a severe accident scenario. To accomplish this task, the vacuum breaker valves and down comers were included in a simplified model of the primary containment that includes both wet well and dry well, which was coupled with a model of the Nuclear Steam Supply System (NSSS), in order to study the behavior of the primary containment during the evolution of the accident scenario. In the analysis of the results of the simulation, the behavior of the wet well and dry well during the event was particularly monitored, by analyzing the evolution of temperature and pressure profiles in such volumes, this to determine the impact of the inclusion of the breaker vacuum valves and down comers. The results show that the effect of this extension of the model is that more conservative results are obtained, i.e., higher pressures are reached in both wet well and dry well than when it is used a containment model that does not include neither the vacuum valves nor the down comers. The most relevant results obtained show that the Rcic alone is able to keep the core fully covered, but even in such a case, it evaporates about 15% of the initial inventory of liquid water in the Pressure Suppression Pool (Psp). When the Rcic operation is lost, 20% more of the liquid water inventory in the Psp is further reduced within four to twelve hours (approximately), time at which the simulation crashed. Besides, there is a significant increase of pressure in the containment. As the accident evolves, the pressure in the containment continues increasing, but there is still considerable margin to reach the design pressure of the containment. At the end of the simulation, the results show a gauge pressure value of 224,550 Pa in the Psp and 187,482 Pa in the wet well

  13. High integrity container evaluation for solid waste disposal burial containers

    International Nuclear Information System (INIS)

    Josephson, W.S.

    1996-01-01

    In order to provide radioactive waste disposal practices with the greatest measure of public protection, Solid Waste Disposal (SWD) adopted the Nuclear Regulatory Commission (NRC) requirement to stabilize high specific activity radioactive waste prior to disposal. Under NRC guidelines, stability may be provided by several mechanisms, one of which is by placing the waste in a high integrity container (HIC). During the implementation process, SWD found that commercially-available HICs could not accommodate the varied nature of weapons complex waste, and in response developed a number of disposal containers to function as HICs. This document summarizes the evaluation of various containers that can be used for the disposal of Category 3 waste in the Low Level Burial Grounds. These containers include the VECTRA reinforced concrete HIC, reinforced concrete culvert, and the reinforced concrete vault. This evaluation provides justification for the use of these containers and identifies the conditions for use of each

  14. Indicator Amino Acid-Derived Estimate of Dietary Protein Requirement for Male Bodybuilders on a Nontraining Day Is Several-Fold Greater than the Current Recommended Dietary Allowance.

    Science.gov (United States)

    Bandegan, Arash; Courtney-Martin, Glenda; Rafii, Mahroukh; Pencharz, Paul B; Lemon, Peter Wr

    2017-05-01

    Background: Despite a number of studies indicating increased dietary protein needs in bodybuilders with the use of the nitrogen balance technique, the Institute of Medicine (2005) has concluded, based in part on methodologic concerns, that "no additional dietary protein is suggested for healthy adults undertaking resistance or endurance exercise." Objective: The aim of the study was to assess the dietary protein requirement of healthy young male bodybuilders ( with ≥3 y training experience) on a nontraining day by measuring the oxidation of ingested l-[1- 13 C]phenylalanine to 13 CO 2 in response to graded intakes of protein [indicator amino acid oxidation (IAAO) technique]. Methods: Eight men (means ± SDs: age, 22.5 ± 1.7 y; weight, 83.9 ± 11.6 kg; 13.0% ± 6.3% body fat) were studied at rest on a nontraining day, on several occasions (4-8 times) each with protein intakes ranging from 0.1 to 3.5 g · kg -1 · d -1 , for a total of 42 experiments. The diets provided energy at 1.5 times each individual's measured resting energy expenditure and were isoenergetic across all treatments. Protein was fed as an amino acid mixture based on the protein pattern in egg, except for phenylalanine and tyrosine, which were maintained at constant amounts across all protein intakes. For 2 d before the study, all participants consumed 1.5 g protein · kg -1 · d -1 On the study day, the protein requirement was determined by identifying the breakpoint in the F 13 CO 2 with graded amounts of dietary protein [mixed-effects change-point regression analysis of F 13 CO 2 (labeled tracer oxidation in breath)]. Results: The Estimated Average Requirement (EAR) of protein and the upper 95% CI RDA for these young male bodybuilders were 1.7 and 2.2 g · kg -1 · d -1 , respectively. Conclusion: These IAAO data suggest that the protein EAR and recommended intake for male bodybuilders at rest on a nontraining day exceed the current recommendations of the Institute of Medicine by ∼2.6-fold

  15. Accident resistant transport container

    Science.gov (United States)

    Anderson, J.A.; Cole, K.K.

    The invention relates to a container for the safe air transport of plutonium having several intermediate wood layers and a load spreader intermediate an inner container and an outer shell for mitigation of shock during a hypothetical accident.

  16. Accident resistant transport container

    International Nuclear Information System (INIS)

    Andersen, J.A.; Cole, J.K.

    1980-01-01

    The invention relates to a container for the safe air transport of plutonium having several intermediate wood layers and a load spreader intermediate an inner container and an outer shell for mitigation of shock during a hypothetical accident

  17. Effectiveness of containment sprays in containment management

    International Nuclear Information System (INIS)

    Nourbakhsh, H.P.; Perez, S.E.; Lehner, J.R.

    1993-05-01

    A limited study has been performed assessing the effectiveness of containment sprays-to mitigate particular challenges which may occur during a severe accident. Certain aspects of three specific topics related to using sprays under severe accident conditions were investigated. The first was the effectiveness of sprays connected to an alternate water supple and pumping source because the actual containment spray pumps are inoperable. This situation could occur during a station blackout. The second topic concerned the adverse as well as beneficial effects of using containment sprays during severe accident scenario where the containment atmosphere contains substantial quantities of hydrogen along with steam. The third topic was the feasibility of using containment sprays to moderate the consequences of DCH

  18. Containment long-term operational integrity

    International Nuclear Information System (INIS)

    Sammataro, R.F.

    1990-01-01

    Periodic integrated leak rate tests are required to assure that containments continue to meet allowable leakage limits. Although overall performance has been quite good to date, several major containment aging and degradation mechanisms have been identified. Two pilot plant life extension (PLEX) studies serve as models for extending the operational integrity of present containments for light-water cooled nuclear power plants in the United States. One study is for a Boiling-Water Reactor (BWR) and the second is for a Pressurized-Water Reactor (PWR). Research and testing programs for determining the ultimate pressure capacity and failure mechanisms for containments under severe loading conditions and studies for extending the life of current plants beyond the present 40-year licensed lifetime are under way. This paper presents an overview of containment designs in the United States. Also presented are a discussion of the American Society of Mechanical Engineers Boiler and Pressure Vessel Code (ASME Code) and regulatory authority requirements for the design, construction, inservice inspection, leakage testing and repair of steel and concrete containments. Findings for containments from the pilot PLEX studies and continuing containment integrity research and testing programs are discussed. The ASME Code and regulatory requirements together with recommendations from the PLEX studies and containment integrity research and testing provide a basis for continued containment long-term operational integrity. (orig./GL)

  19. Use of Oritavancin in Moderate-to-Severe ABSSSI Patients Requiring IV Antibiotics: A U.S. Payer Budget Impact Analysis.

    Science.gov (United States)

    Jensen, Ivar S; Wu, Elizabeth; Fan, Weihong; Lodise, Thomas P; Nicolau, David P; Dufour, Scott; Cyr, Philip L; Sulham, Katherine A

    2016-06-01

    It is estimated that acute bacterial skin and skin structure infections (ABSSSI) account for nearly 10% of hospital admissions and 3.4-3.8 million emergency department visits per year in the United States. Analyses of hospital discharge records indicate 74% of ABSSSI admissions involve empiric treatment with methicillin-resistant Staphylococcus aureus (MRSA) active antibiotics. Analysis has shown that payer costs could be reduced if moderate-to-severe ABSSSI patients were treated to a greater extent in the observational unit followed by discharge to outpatient parenteral antibiotic therapy (OPAT). Oritavancin is a lipoglycopeptide antibiotic with bactericidal activity against gram-positive bacteria, including MRSA. To estimate the impact on a U.S. payer's budget of using single-dose oritavancin in ABSSSI patients with suspected MRSA involvement who are indicated for intravenous antibiotics. A decision analytic model based on current clinical practice was developed to estimate the economic value of decreased hospital resource consumption by using single-dose oritavancin over a 1-year time horizon. Use of antibiotics was informed by an analysis of the Premier Research Database. Demographic and clinical data were derived from a targeted literature review. Emergency department, observation, laboratory, and administration costs used were Medicare National Limitation amounts. Drug costs were 2014 wholesale acquisition costs. For a hypothetical U.S. payer with 1,000,000 members, it is expected that approximately 14,285 members per year will be diagnosed with ABSSSI severe enough to indicate intravenous antibiotics with MRSA activity. Based on this simulation, use of single-dose oritavancin in 26% of these patients was estimated to reduce the number of inpatient admissions, reduce length of stay for patients requiring admission, and reduce the number of days a patient needs to receive daily infusions in the OPAT clinic. The total patient days decreased from 171,125 to 133

  20. Containment performance improvement program

    International Nuclear Information System (INIS)

    Beckner, W.; Mitchell, J.; Soffer, L.; Chow, E.; Lane, J.; Ridgely, J.

    1990-01-01

    The Containment Performance Improvement (CPI) program has been one of the main elements in the US Nuclear Regulatory Commission's (NRC's) integrated approach to closure of severe accident issues for US nuclear power plants. During the course of the program, results from various probabilistic risk assessment (PRA) studies and from severe accident research programs for the five US containment types have been examined to identify significant containment challenges and to evaluate potential improvements. The five containment types considered are: the boiling water reactor (BMR) Mark I containment, the BWR Mark II containment, the BWR Mark III containment, the pressurized water reactor (PWR) ice condenser containment, and the PWR dry containments (including both subatmospheric and large subtypes). The focus of the CPI program has been containment performance and accident mitigation, however, insights are also being obtained in the areas of accident prevention and accident management

  1. Development of a metallic high integrity container

    International Nuclear Information System (INIS)

    Haelsig, R.T.

    1986-01-01

    Nuclear Packaging, a Pacific Nuclear Company, developed a metallic high integrity container (HIC) for the burial of low level radioactive waste. This class of container has received the most extensive review of any burial container licensed in the United States. It is also the first container that has been licensed to meet the requirements of Nuclear Regulatory Commission Regulations 10CFR61. The design and subsequent review considered 300 years corrosion at a depth of 55 feet with no degradation of container structural integrity. The design also included a technical requirement that the container possess a positive vent that would exclude moisture. The alloy that was selected, allows for significant flexibility in container size and configuration which is essential to accommodating the various waste forms. This allowed the development of containers in various sizes and with a variety of closures, that accommodate the internal dimensions of various shipping shields and help minimize radiation exposure during packaging operations. The material used in the metallic container is high corrosion resistant which reduces the need for strict chemical controls at the waste generating facility. This acts to ease the operational requirements in the treatment of several waste streams. The design result is a family of metallic High Integrity Containers (HIC)s that meet all the performance criteria imposed by the regulations, as well as provide a disposable waste container with good transportation efficiency and minimum operational constraints

  2. FAUST/CONTAIN; FAUST/CONTAIN

    Energy Technology Data Exchange (ETDEWEB)

    Cherdron, W.; Minges, J.; Sauter, H.; Schuetz, W.

    1995-08-01

    The FAUNA facility has been restructured after completion of the sodium fire experiments. It is now serving LWR research, cf. report II on program no. 32.21.02 concerning steam explosions. The CONTAIN code system for computing the thermodynamic, aerosol and radiological phenomena in a containment under severe accident conditions is being developed with a new to fission product release and transport. (orig.)

  3. SPHINGOLIPID-DEPENDENT FUSION OF SEMLIKI FOREST VIRUS WITH CHOLESTEROL-CONTAINING LIPOSOMES REQUIRES BOTH THE 3-HYDROXYL GROUP AND THE DOUBLE-BOND OF THE SPHINGOLIPID BACKBONE

    NARCIS (Netherlands)

    CORVER, J; MOESBY, L; ERUKULLA, RK; REDDY, KC; BITTMAN, R; WILSCHUT, J

    Low-pH-induced membrane fusion of Semliki Forest virus (SFV) in a model system is mediated by sphingolipids in the target membrane; ceramide is the sphingolipid minimally required (J. L. Nieva, R. Bron, J. Corver, and J. Wilschut, EMBO J. 13:2797-2804, 1994). Here, using various ceramide analogs, we

  4. Surface motility in Pseudomonas sp DSS73 is required for efficient biological containment of the root-pathogenic microfungi Rhizoctonia solani and Pythium ultimum

    DEFF Research Database (Denmark)

    Andersen, Jens Bo; Koch, Birgit; Nielsen, T.H.

    2003-01-01

    Pseudomonas sp. DSS73 was isolated from the rhizoplane of sugar beet seedlings. This strain exhibits antagonism towards the root-pathogenic microfungi Pythium ultimum and Rhizoctonia solani. Production of the cyclic lipopeptide amphisin in combination with expression of flagella enables the growing......-pathogenic microfungi is shown to arise from amphisin-dependent surface translocation and growth by which the bacterium can lay siege to the fungi. The synergistic effects of surface motility and synthesis of a battery of antifungal compounds efficiently contain and terminate growth of the microfungi....

  5. CANDU safety under severe accidents

    International Nuclear Information System (INIS)

    Snell, V.G.; Howieson, J.Q.; Alikhan, S.; Frescura, G.M.; King, F.; Rogers, J.T.; Tamm, H.

    1996-01-01

    The characteristics of the CANDU reactor relevant to severe accidents are set first by the inherent properties of the design, and second by the Canadian safety/licensing approach. The pressure-tube concept allows the separate, low-pressure, heavy-water moderator to act as a backup heat sink even if there is no water in the fuel channels. Should this also fail, the calandria shell itself can contain the debris, with heat being transferred to the water-filled shield tank around the core. Should the severe core damage sequence progress further, the shield tank and the concrete reactor vault significantly delay the challenge to containment. Furthermore, should core melt lead to containment overpressure, the containment behaviour is such that leaks through the concrete containment wall reduce the possibility of catastrophic structural failure. The Canadian licensing philosophy requires that each accident, together with failure of each safety system in turn, be assessed (and specified dose limits met) as part of the design and licensing basis. In response, designers have provided CANDUs with two independent dedicated shutdown systems, and the likelihood of Anticipated Transients Without Scram is negligible. Probabilistic safety assessment studies have been performed on operating CANDU plants, and on the 4 x 880 MW(e) Darlington station now under construction; furthermore a scoping risk assessment has been done for a CANDU 600 plant. They indicate that the summed severe core damage frequency is of the order of 5 x 10 -6 /year. 95 refs, 3 tabs

  6. CANDU safety under severe accidents

    Energy Technology Data Exchange (ETDEWEB)

    Snell, V G; Howieson, J Q [Atomic Energy of Canada Ltd. (Canada); Alikhan, S [New Brunswick Electric Power Commission (Canada); Frescura, G M; King, F [Ontario Hydro (Canada); Rogers, J T [Carleton Univ., Ottawa, ON (Canada); Tamm, H [Atomic Energy of Canada Ltd. (Canada). Whiteshell Research Lab.

    1996-12-01

    The characteristics of the CANDU reactor relevant to severe accidents are set first by the inherent properties of the design, and second by the Canadian safety/licensing approach. The pressure-tube concept allows the separate, low-pressure, heavy-water moderator to act as a backup heat sink even if there is no water in the fuel channels. Should this also fail, the calandria shell itself can contain the debris, with heat being transferred to the water-filled shield tank around the core. Should the severe core damage sequence progress further, the shield tank and the concrete reactor vault significantly delay the challenge to containment. Furthermore, should core melt lead to containment overpressure, the containment behaviour is such that leaks through the concrete containment wall reduce the possibility of catastrophic structural failure. The Canadian licensing philosophy requires that each accident, together with failure of each safety system in turn, be assessed (and specified dose limits met) as part of the design and licensing basis. In response, designers have provided CANDUs with two independent dedicated shutdown systems, and the likelihood of Anticipated Transients Without Scram is negligible. Probabilistic safety assessment studies have been performed on operating CANDU plants, and on the 4 x 880 MW(e) Darlington station now under construction; furthermore a scoping risk assessment has been done for a CANDU 600 plant. They indicate that the summed severe core damage frequency is of the order of 5 x 10{sup -6}/year. 95 refs, 3 tabs.

  7. Containment structure tendon investigation

    International Nuclear Information System (INIS)

    Fulton, J.F.; Murray, K.H.

    1983-01-01

    The paper describes an investigation into the possible causes of lower-than-predicted tendon forces which were measured during past tendon surveillances for a concrete containment. The containment is post tensioned by vertical tendons which are anchored into a rock foundation. The tendons were originally stressed in 1969, and lift-off tests were performed on six occasions subsequent to this date over a period of 11 years. The tendon forces measured in these tests were generally lower than predicted, and by 1979 the prestress level in the containment was only marginally above the design requirement. The tendons were retensioned in 1980, and by this time an investigation into the possible causes was underway. Potential causes investigated include the rock anchors and surrounding rock, elastomeric pad creep, wire stresses, thermal effects, stressing equipment and lift-off procedures, and wire stress relaxation. The investigation activities included stress relaxation testing of wires pulled from actual tendons. The stress relaxation test program included wire specimens at several different temperature and initial stress levels and the effect of a varying temperature history on the stress relaxation property of the wires. For purpose of future force predictions of the retensioned tendons, the test program included tests to determine the effect on stress relaxation due to restressing the wires after they had relaxed for 1000 hours and 10,000 hours. (orig./GL)

  8. Angiotensin-converting enzyme and angiotensin II receptor subtype 2 genotypes in type 1 diabetes and severe hypoglycaemia requiring emergency treatment: a case cohort study

    DEFF Research Database (Denmark)

    Pedersen-Bjergaard, Ulrik; Nielsen, Søren L; Akram, Kamran

    2009-01-01

    AIMS: In type 1 diabetes, individual susceptibility to severe hypoglycaemia is likely to be influenced by genetic factors. We have previously reported an association of the deletion (D-) allele of the angiotensin-converting enzyme (ACE) insertion/deletion (I/D) polymorphism and the A-allele of th...

  9. Proceedings of the 2. international conference on containment design and operation. Vol. 1,2

    Energy Technology Data Exchange (ETDEWEB)

    Lawrence, S R

    1991-12-31

    The second international conference on containment design and operation included sessions on the following topics: plenary; commissioning and operation; regulatory and performance requirements; reliability, risk and severe accident evaluation; future containment systems; severe accident evaluation; activity transport experiments; activity transport analysis; containment design; future and filtered vent containment systems; containment response; hydrogen combustion. Due prominence was given to CANDU and other PHWR reactors, and to Canadian experiments. The individual papers have been abstracted separately.

  10. Several tetratricopeptide repeat (TPR) motifs of FANCG are required for assembly of the BRCA2/D1-D2-G-X3 complex, FANCD2 monoubiquitylation and phleomycin resistance

    International Nuclear Information System (INIS)

    Wilson, James B.; Blom, Eric; Cunningham, Ryan; Xiao, Yuxuan; Kupfer, Gary M.; Jones, Nigel J.

    2010-01-01

    The Fanconi anaemia (FA) FANCG protein is an integral component of the FA nuclear core complex that is required for monoubiquitylation of FANCD2. FANCG is also part of another protein complex termed D1-D2-G-X3 that contains FANCD2 and the homologous recombination repair proteins BRCA2 (FANCD1) and XRCC3. Formation of the D1-D2-G-X3 complex is mediated by serine-7 phosphorylation of FANCG and occurs independently of the FA core complex and FANCD2 monoubiquitylation. FANCG contains seven tetratricopeptide repeat (TPR) motifs that mediate protein-protein interactions and here we show that mutation of several of the TPR motifs at a conserved consensus residue ablates the in vivo binding activity of FANCG. Expression of mutated TPR1, TPR2, TPR5 and TPR6 in Chinese hamster fancg mutant NM3 fails to functionally complement its hypersensitivities to mitomycin C (MMC) and phleomycin and fails to restore FANCD2 monoubiquitylation. Using co-immunoprecipitation analysis, we demonstrate that these TPR-mutated FANCG proteins fail to interact with BRCA2, XRCC3, FANCA or FANCF. The interactions of other proteins in the D1-D2-G-X3 complex are also absent, including the interaction of BRCA2 with both the monoubiquitylated (FANCD2-L) and non-ubiquitylated (FANCD2-S) isoforms of FANCD2. Interestingly, a mutation of TPR7 (R563E), that complements the MMC and phleomycin hypersensitivity of human FA-G EUFA316 cells, fails to complement NM3, despite the mutated FANCG protein co-precipitating with FANCA, BRCA2 and XRCC3. Whilst interaction of TPR7-mutated FANCG with FANCF does appear to be reduced in NM3, FANCD2 is monoubiquitylated suggesting that sub-optimal interactions of FANCG in the core complex and the D1-D2-G-X3 complex are responsible for the observed MMC- and phleomycin-hypersensitivity, rather than a defect in FANCD2 monoubiquitylation. Our data demonstrate that FANCG functions as a mediator of protein-protein interactions and is vital for the assembly of multi-protein complexes

  11. Several tetratricopeptide repeat (TPR) motifs of FANCG are required for assembly of the BRCA2/D1-D2-G-X3 complex, FANCD2 monoubiquitylation and phleomycin resistance

    Energy Technology Data Exchange (ETDEWEB)

    Wilson, James B. [Molecular Oncology and Stem Cell Research Group, School of Biological Sciences, University of Liverpool, Biosciences Building, Crown Street, Liverpool L69 7ZB (United Kingdom); Blom, Eric [Department of Clinical Genetics and Human Genetics, VU University Medical Center, Van der Boechorststraat 7, NL-1081 BT Amsterdam (Netherlands); Cunningham, Ryan; Xiao, Yuxuan [Molecular Oncology and Stem Cell Research Group, School of Biological Sciences, University of Liverpool, Biosciences Building, Crown Street, Liverpool L69 7ZB (United Kingdom); Kupfer, Gary M. [Departments of Pediatrics and Pathology, Yale University School of Medicine, Section of Hematology/Oncology, 333 Cedar Street, New Haven, CT 0652 (United States); Jones, Nigel J., E-mail: njjones@liv.ac.uk [Molecular Oncology and Stem Cell Research Group, School of Biological Sciences, University of Liverpool, Biosciences Building, Crown Street, Liverpool L69 7ZB (United Kingdom)

    2010-07-07

    The Fanconi anaemia (FA) FANCG protein is an integral component of the FA nuclear core complex that is required for monoubiquitylation of FANCD2. FANCG is also part of another protein complex termed D1-D2-G-X3 that contains FANCD2 and the homologous recombination repair proteins BRCA2 (FANCD1) and XRCC3. Formation of the D1-D2-G-X3 complex is mediated by serine-7 phosphorylation of FANCG and occurs independently of the FA core complex and FANCD2 monoubiquitylation. FANCG contains seven tetratricopeptide repeat (TPR) motifs that mediate protein-protein interactions and here we show that mutation of several of the TPR motifs at a conserved consensus residue ablates the in vivo binding activity of FANCG. Expression of mutated TPR1, TPR2, TPR5 and TPR6 in Chinese hamster fancg mutant NM3 fails to functionally complement its hypersensitivities to mitomycin C (MMC) and phleomycin and fails to restore FANCD2 monoubiquitylation. Using co-immunoprecipitation analysis, we demonstrate that these TPR-mutated FANCG proteins fail to interact with BRCA2, XRCC3, FANCA or FANCF. The interactions of other proteins in the D1-D2-G-X3 complex are also absent, including the interaction of BRCA2 with both the monoubiquitylated (FANCD2-L) and non-ubiquitylated (FANCD2-S) isoforms of FANCD2. Interestingly, a mutation of TPR7 (R563E), that complements the MMC and phleomycin hypersensitivity of human FA-G EUFA316 cells, fails to complement NM3, despite the mutated FANCG protein co-precipitating with FANCA, BRCA2 and XRCC3. Whilst interaction of TPR7-mutated FANCG with FANCF does appear to be reduced in NM3, FANCD2 is monoubiquitylated suggesting that sub-optimal interactions of FANCG in the core complex and the D1-D2-G-X3 complex are responsible for the observed MMC- and phleomycin-hypersensitivity, rather than a defect in FANCD2 monoubiquitylation. Our data demonstrate that FANCG functions as a mediator of protein-protein interactions and is vital for the assembly of multi-protein complexes

  12. The Genome of a Tortoise Herpesvirus (Testudinid Herpesvirus 3) Has a Novel Structure and Contains a Large Region That Is Not Required for Replication In Vitro or Virulence In Vivo

    Science.gov (United States)

    Gandar, Frédéric; Wilkie, Gavin S.; Gatherer, Derek; Kerr, Karen; Marlier, Didier; Diez, Marianne; Marschang, Rachel E.; Mast, Jan; Dewals, Benjamin G.

    2015-01-01

    vitro, and investigated the pathogenesis of strain 4295, which consists of three deletion mutants. The major findings are that (i) TeHV-3 has a novel genome structure, (ii) its closest relative is a turtle herpesvirus, (iii) it contains interleukin-10 and semaphorin genes (the first time these have been reported in an alphaherpesvirus), (iv) a sizeable region of the genome is not required for viral replication in vitro or virulence in vivo, and (v) one of the components of strain 4295, which has a deletion of 22.4 kb, exhibits properties indicating that it may serve as the starting point for an attenuated vaccine. PMID:26339050

  13. Mass spectrometry and site-directed mutagenesis identify several autophosphorylated residues required for the activity of PrkC, a Ser/Thr kinase from Bacillus subtilis

    DEFF Research Database (Denmark)

    Madec, Edwige; Stensballe, Allan; Kjellström, Sven

    2003-01-01

    We have shown recently that PrkC, which is involved in developmental processes in Bacillus subtilis, is a Ser/Thr kinase with features of the receptor kinase family of eukaryotic Hanks kinases. In this study, we expressed and purified from Escherichia coli the cytoplasmic domain of PrkC containing...... the kinase and a short juxtamembrane region. This fragment, which we designate PrkCc, undergoes autophosphorylation in E.coli. PrkCc is further autophosphorylated in vitro, apparently through a trans-kinase, intermolecular reaction. PrkC also displays kinase activity with myelin basic protein. Using high...... mass accuracy electrospray tandem mass spectrometry (LC-MS/MS) and nanoelectrospray tandem mass spectrometry, we identified seven phosphorylated threonine and one serine residue in PrkCc. All the corresponding residues were replaced by systematic site-directed mutagenesis and the purified mutant...

  14. Shielding container

    International Nuclear Information System (INIS)

    Darling, K.A.M.

    1981-01-01

    A shielding container incorporates a dense shield, for example of depleted uranium, cast around a tubular member of curvilinear configuration for accommodating a radiation source capsule. A lining for the tubular member, in the form of a close-coiled flexible guide, provides easy replaceability to counter wear while the container is in service. Container life is extended, and maintenance costs are reduced. (author)

  15. Severe accident phenomena

    International Nuclear Information System (INIS)

    Jokiniemi, J.; Kilpi, K.; Lindholm, I.; Maekynen, J.; Pekkarinen, E.; Sairanen, R.; Silde, A.

    1995-02-01

    Severe accidents are nuclear reactor accidents in which the reactor core is substantially damaged. The report describes severe reactor accident phenomena and their significance for the safety of nuclear power plants. A comprehensive set of phenomena ranging from accident initiation to containment behaviour and containment integrity questions are covered. The report is based on expertise gained in the severe accident assessment projects conducted at the Technical Research Centre of Finland (VTT). (49 refs., 32 figs., 12 tabs.)

  16. Long term integrity of reactor pressure vessel and primary containment vessel after the severe accidents in Fukushima Daiichi Nuclear Power Station. Leaching property of spent oxide fuel segment and corrosion property of a carbon steel under artificial seawater immersion

    International Nuclear Information System (INIS)

    2014-06-01

    Primary containment vessel (PCV), reactor pressure vessel and pedestal in Fukushima Daiichi Nuclear power station units 1 through 3 have been exposed to severe thermal, chemical and mechanical conditions due to core meltdown events and seawater injections for emergent core cooling. These components will be immersed in diluted seawater with dissolved fission products under irradiation until the end of debris removal. Fresh water injected into the cores contacts with debris to cool, dissolves or erodes their constituents, mixed with retained water, and becomes 'accumulated water' with radioactive nuclides. We have focused the leaching of fission products into the accumulated water under lower temperature (323 K). FUGEN spent oxide fuel segments were immersed to determine the leaching factor of fission product and actinide elements. Since PCV made from carbon steel is one of the most important boundaries to prevent from fission products release, corrosion behavior has been paid attention to evaluate their integrity. Carbon steel specimens were immersion- and electrochemical-tested in diluted seawater with simulants of the accumulated water at 323 K in order to evaluate the effect of fission products in particular cesium and radiation. (author)

  17. Localization of a sound source in in a guided medium and reverberating field. Contribution to a study on leak localization in the internal wall of containment of a nuclear reactor in the case of a severe reactor accident

    International Nuclear Information System (INIS)

    Thomann, F.

    1996-01-01

    Basic data necessary for the localization of a leak in the internal wall of the containment are presented by studying the sound generated by gas jets coming out of (leaking fissures) as well as propagation in a guided medium. The results acquired have led us to choose the simple intercorrelation method and the matched filed processing method, both of which are likely to adequately handle our problems. Whereas the intercorrelation method appears to be limited in scope when dealing in the guided medium, the matched field processing is suited to leak localization over a surface of approximately 1000 m 2 (for a total surface of 10 000 m 2 ). Preliminary studies on the leak signal and on replicated signals have led us to limit the frequency band to 2600 - 3000 Hz. We have succeeded in locating a leak situated in an ordinary position with a minimum amount of replicated signals and basic data. We have improved on the estimation of Bartlett and MVDE (minimum variance distortion less filter) rendering them even more effective. Afterwards, we considered the severe accident situation and showed that the system can be installed in situ. (author)

  18. Corrosion Study of Super Ferritic Stainless Steel UNS S44660 (26Cr-3Ni-3Mo) and Several Other Stainless Steel Grades (UNS S31603, S32101, and S32205) in Caustic Solution Containing Sodium Sulfide

    Science.gov (United States)

    Chasse, Kevin R.; Singh, Preet M.

    2013-11-01

    Electrochemical techniques, scanning electron microscopy (SEM), and X-ray photoelectron spectroscopy (XPS) were used in this study to show how the corrosion mechanism of several commercial grades of stainless steel in hot caustic solution is strongly influenced by the presence of sodium sulfide. Experimental results from super ferritic stainless steel UNS S44660 (26Cr-3Ni-3Mo) were compared to austenitic stainless steel UNS S31603, lean duplex stainless steel (DSS) UNS S32101, and standard DSS UNS S32205 in caustic solution, with and without sodium sulfide, at 443 K (170 °C). Weight loss measurements indicated that corrosion rates of UNS44660 were much lower than the other grades of stainless steel in the presence of the sodium sulfide. Potentiodynamic polarization and linear polarization resistance measurements showed that the electrochemical behavior was altered by the adhesion of sulfur species, which reduced the polarization resistances and increased the anodic current densities. SEM and XPS results imply that the surface films that formed in caustic solution containing sodium sulfide were defective due to the adsorption of sulfide, which destabilized the passive film and led to the formation of insoluble metal sulfide compounds.

  19. Several requirements made on the quality of nuclide-specific gamma radiation measuring rigs with germanium detectors for nuclear power plant monitoring

    International Nuclear Information System (INIS)

    Wuensch, K.D.

    1986-01-01

    The measuring set-ups are used for nuclide-specific analysis of all samples from nuclear power plants, i.e. solid, liquid, or gaseous. Depending on the nature of the sample, various requirements made by ordinances and guidelines discussed in detail must be met, relating to emission monitoring, environmental monitoring, system monitoring, room air monitoring, and contamination monitoring. The state-of-the-art is shown emphasizing the resolution and the evaluation scheme. Experience gained in the control of such systems is reported on in brief. The quality of in-service inspections is discussed. (orig./PW) [de

  20. 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.)

  1. The cholesterol, fatty acid and triglyceride synthesis pathways regulated by site 1 protease (S1P) are required for efficient replication of severe fever with thrombocytopenia syndrome virus.

    Science.gov (United States)

    Urata, Shuzo; Uno, Yukiko; Kurosaki, Yohei; Yasuda, Jiro

    2018-06-12

    Severe fever with thrombocytopenia syndrome (SFTS) is an emerging infectious disease caused by the SFTS virus (SFTSV), which has a high mortality rate. Currently, no licensed vaccines or therapeutic agents have been approved for use against SFTSV infection. Here, we report that the cholesterol, fatty acid, and triglyceride synthesis pathways regulated by S1P is involved in SFTSV replication, using CHO-K1 cell line (SRD-12B) that is deficient in site 1 protease (S1P) enzymatic activity, PF-429242, a small compound targeting S1P enzymatic activity, and Fenofibrate and Lovastatin, which inhibit triglyceride and cholesterol synthesis, respectively. These results enhance our understanding of the SFTSV replication mechanism and may contribute to the development of novel therapies for SFTSV infection. Copyright © 2018. Published by Elsevier Inc.

  2. Exemption from licensing requirements for spark-gap irradiators that contain cobalt-60: (Docket No. PRM 30-54): Draft environmental statement

    International Nuclear Information System (INIS)

    1975-09-01

    This Draft Statement on environmental considerations associated with the proposed exemption of cobalt-60 in spark-gap irradiators was prepared by the US Nuclear Regulatory Commission, Office of Standards Development (the staff), in accordance with the Commission's regulation 10 CFR Part 51, which implements the requirements of the National Environmental Policy Act of 1969 (NEPA). The NEPA states, among other things, that the federal government has the continuing responsibility to use all practicable means, consistent with other essential considerations of national policy, to improve and to coordinate federal plans, functions, programs, and resources to the end that the nation may: fulfill the responsibilities of each generation as trustee of the environment for succeeding generations; assure for all Americans safe, healthful, productive, and esthetically and culturally pleasing surroundings; attain the widest range of beneficial uses of the environment without degradation, risk to health or safety, or other undesirable and unintended consequences; preserve important historic, cultural, and natural aspects of our national heritage, and maintain, wherever possible, an environment which supports diversity and variety of individual choice; achieve a balance between population and resource use which will permit high standards of living and a wide sharing of life's amenities; and enhance the quality of renewable resources and approach the maximum attainable recycling of depletable resources. 42 refs., 5 figs., 7 tabs

  3. Infection with CagA-positive Helicobacter pylori strain containing three EPIYA C phosphorylation sites is associated with more severe gastric lesions in experimentally infected Mongolian gerbils (Meriones unguiculatus).

    Science.gov (United States)

    Ferreira Júnior, M; Batista, S A; Vidigal, P V T; Cordeiro, A A C; Oliveira, F M S; Prata, L O; Diniz, A E T; Barral, C M; Barbuto, R C; Gomes, A D; Araújo, I D; Queiroz, D M M; Caliari, M V

    2015-04-27

    Infection with Helicobacter pylori strains containing high number of EPIYA-C phosphorylation sites in the CagA is associated with significant gastritis and increased risk of developing pre-malignant gastric lesions and gastric carcinoma. However, these findings have not been reproduced in animal models yet. Therefore, we investigated the effect on the gastric mucosa of Mongolian gerbil (Meriones unguiculatus) infected with CagA-positive H. pylori strains exhibiting one or three EPIYA-C phosphorilation sites. Mongolian gerbils were inoculated with H. pylori clonal isolates containing one or three EPIYA-C phosphorylation sites. Control group was composed by uninfected animals challenged with Brucella broth alone. Gastric fragments were evaluated by the modified Sydney System and digital morphometry. Clonal relatedness between the isolates was considered by the identical RAPD-PCR profiles and sequencing of five housekeeping genes, vacA i/d region and of oipA. The other virulence markers were present in both isolates (vacA s1i1d1m1, iceA2, and intact dupA). CagA of both isolates was translocated and phosphorylated in AGS cells. After 45 days of infection, there was a significant increase in the number of inflammatory cells and in the area of the lamina propria in the infected animals, notably in those infected by the CagA-positive strain with three EPIYA-C phosphorylation sites. After six months of infection, a high number of EPIYA-C phosphorylation sites was associated with progressive increase in the intensity of gastritis and in the area of the lamina propria. Atrophy, intestinal metaplasia, and dysplasia were also observed more frequently in animals infected with the CagA-positive isolate with three EPIYA-C sites.  We conclude that infection with H. pylori strain carrying a high number of CagA EPIYA-C phosphorylation sites is associated with more severe gastric lesions in an animal model of H. pylori infection.

  4. Infection with CagA-positive Helicobacter pylori strain containing three EPIYA C phosphorylation sites is associated with more severe gastric lesions in experimentally infected Mongolian gerbils (Meriones unguiculatus

    Directory of Open Access Journals (Sweden)

    M. Ferreira Júnior

    2015-04-01

    Full Text Available Infection with Helicobacter pylori strains containing high number of EPIYA-C phosphorylation sites in the CagA is associated with significant gastritis and increased risk of developing pre-malignant gastric lesions and gastric carcinoma. However, these findings have not been reproduced in animal models yet. Therefore, we investigated the effect on the gastric mucosa of Mongolian gerbil (Meriones unguiculatus infected with CagA-positive H. pylori strains exhibiting one or three EPIYA-C phosphorilation sites. Mongolian gerbils were inoculated with H. pylori clonal isolates containing one or three EPIYA-C phosphorylation sites. Control group was composed by uninfected animals challenged with Brucella broth alone. Gastric fragments were evaluated by the modified Sydney System and digital morphometry. Clonal relatedness between the isolates was considered by the identical RAPD-PCR profiles and sequencing of five housekeeping genes, vacA i/d region and of oipA. The other virulence markers were present in both isolates (vacA s1i1d1m1, iceA2, and intact dupA. CagA of both isolates was translocated and phosphorylated in AGS cells. After 45 days of infection, there was a significant increase in the number of inflammatory cells and in the area of the lamina propria in the infected animals, notably in those infected by the CagA-positive strain with three EPIYA-C phosphorylation sites. After six months of infection, a high number of EPIYA-C phosphorylation sites was associated with progressive increase in the intensity of gastritis and in the area of the lamina propria. Atrophy, intestinal metaplasia, and dysplasia were also observed more frequently in animals infected with the CagA-positive isolate with three EPIYA-C sites.  We conclude that infection with H. pylori strain carrying a high number of CagA EPIYA-C phosphorylation sites is associated with more severe gastric lesions in an animal model of H. pylori infection.

  5. Evidence for a Proton Transfer Network and a Required Persulfide-Bond-Forming Cysteine Residue in Ni-Containing Carbon Monoxide Dehydrogenases

    International Nuclear Information System (INIS)

    Eun Jin Kim; Jian Feng; Bramlett, Matthew R.; Lindahl, Paul A.

    2004-01-01

    OAK-B135 Carbon monoxide dehydrogenase from Moorella thermoacetica catalyzes the reversible oxidation of CO to CO2 at a nickel-iron-sulfur active-site called the C-cluster. Mutants of a proposed proton transfer pathway and of a cysteine residue recently found to form a persulfide bond with the C-cluster were characterized. Four semi-conserved histidine residues were individually mutated to alanine. His116 and His122 were essential to catalysis, while His113 and His119 attenuated catalysis but were not essential. Significant activity was ''rescued'' by a double mutant where His116 was replaced by Ala and His was also introduced at position 115. Activity was also rescued in double mutants where His122 was replaced by Ala and His was simultaneously introduced at either position 121 or 123. Activity was also ''rescued'' by replacing His with Cys at position 116. Mutation of conserved Lys587 near the C-cluster attenuated activity but did not eliminate it. Activity was virtually abolished in a double mutant where Lys587 and His113 were both changed to Ala. Mutations of conserved Asn284 also attenuated activity. These effects suggest the presence of a network of amino acid residues responsible for proton transfer rather than a single linear pathway. The Ser mutant of the persulfide-forming Cys316 was essentially inactive and displayed no EPR signals originating from the C-cluster. Electronic absorption and metal analysis suggests that the C-cluster is absent in this mutant. The persulfide bond appears to be essential for either the assembly or stability of the C-cluster, and/or for eliciting the redox chemistry of the C-cluster required for catalytic activity

  6. Whirlin and PDZ domain-containing 7 (PDZD7) proteins are both required to form the quaternary protein complex associated with Usher syndrome type 2.

    Science.gov (United States)

    Chen, Qian; Zou, Junhuang; Shen, Zuolian; Zhang, Weiping; Yang, Jun

    2014-12-26

    Usher syndrome (USH) is the leading genetic cause of combined hearing and vision loss. Among the three USH clinical types, type 2 (USH2) occurs most commonly. USH2A, GPR98, and WHRN are three known causative genes of USH2, whereas PDZD7 is a modifier gene found in USH2 patients. The proteins encoded by these four USH genes have been proposed to form a multiprotein complex, the USH2 complex, due to interactions found among some of these proteins in vitro, their colocalization in vivo, and mutual dependence of some of these proteins for their normal in vivo localizations. However, evidence showing the formation of the USH2 complex is missing, and details on how this complex is formed remain elusive. Here, we systematically investigated interactions among the intracellular regions of the four USH proteins using colocalization, yeast two-hybrid, and pull-down assays. We show that multiple domains of the four USH proteins interact among one another. Importantly, both WHRN and PDZD7 are required for the complex formation with USH2A and GPR98. In this USH2 quaternary complex, WHRN prefers to bind to USH2A, whereas PDZD7 prefers to bind to GPR98. Interaction between WHRN and PDZD7 is the bridge between USH2A and GPR98. Additionally, the USH2 quaternary complex has a variable stoichiometry. These findings suggest that a non-obligate, short term, and dynamic USH2 quaternary protein complex may exist in vivo. Our work provides valuable insight into the physiological role of the USH2 complex in vivo and informs possible reconstruction of the USH2 complex for future therapy. © 2014 by The American Society for Biochemistry and Molecular Biology, Inc.

  7. Reactor container

    International Nuclear Information System (INIS)

    Naruse, Yoshihiro.

    1990-01-01

    The thickness of steel shell plates in a reactor container embedded in sand cussions is monitored to recognize the corrosion of the steel shell plates. That is, the reactor pressure vessel is contained in a reactor container shell and the sand cussions are disposed on the lower outside of the reactor container shell to elastically support the shell. A pit is disposed at a position opposing to the sand cussions for measuring the thickness of the reactor container shell plates. The pit is usually closed by a closing member. In the reactor container thus constituted, the closing member can be removed upon periodical inspection to measure the thickness of the shell plates. Accordingly, the corrosion of the steel shell plates can be recognized by the change of the plate thickness. (I.S.)

  8. A container

    DEFF Research Database (Denmark)

    2012-01-01

    A container assembly for the containment of fluids or solids under a pressure different from the ambient pressure comprising a container (2) comprising an opening and an annular sealing, a lid (3) comprising a central portion (5) and engagement means (7) for engaging the annular flange, and sealing...... means (10) wherein the engagement means (7) is adapted, via the sealing means, to seal the opening when the pressure of the container assembly differs from the ambient pressure in such a way that the central portion (5) flexes in the axial direction which leads to a radial tightening of the engagement...... means (7) to the container, wherein the container further comprises locking means (12) that can be positioned so that the central portion is hindered from flexing in at least one direction....

  9. Shielded container

    International Nuclear Information System (INIS)

    Fries, B.A.

    1978-01-01

    A shielded container for transportation of radioactive materials is disclosed in which leakage from the container is minimized due to constructional features including, inter alia, forming the container of a series of telescoping members having sliding fits between adjacent side walls and having at least two of the members including machine sealed lids and at least two of the elements including hand-tightenable caps

  10. Aerosol in the containment

    International Nuclear Information System (INIS)

    Lanza, S.; Mariotti, P.

    1986-01-01

    The US program LACE (LWR Aerosol Containment Experiments), in which Italy participates together with several European countries, Canada and Japan, aims at evaluating by means of a large scale experimental activity at HEDL the retention in the pipings and primary container of the radioactive aerosol released following severe accidents in light water reactors. At the same time these experiences will make available data through which the codes used to analyse the behaviour of the aerosol in the containment and to verify whether by means of the codes of thermohydraulic computation it is possible to evaluate with sufficient accuracy variable influencing the aerosol behaviour, can be validated. This report shows and compares the results obtained by the participants in the LACE program with the aerosol containment codes NAVA 5 and CONTAIN for the pre-test computations of the test LA 1, in which an accident called containment by pass is simulated

  11. Definition of containment failure

    International Nuclear Information System (INIS)

    Cybulskis, P.

    1982-01-01

    Core meltdown accidents of the types considered in probabilistic risk assessments (PRA's) have been predicted to lead to pressures that will challenge the integrity of containment structures. Review of a number of PRA's indicates considerable variation in the predicted probability of containment failure as a function of pressure. Since the results of PRA's are sensitive to the prediction of the occurrence and the timing of containment failure, better understanding of realistic containment capabilities and a more consistent approach to the definition of containment failure pressures are required. Additionally, since the size and location of the failure can also significantly influence the prediction of reactor accident risk, further understanding of likely failure modes is required. The thresholds and modes of containment failure may not be independent

  12. Group 4. Containment

    International Nuclear Information System (INIS)

    McCauley, V.S.; Keiser, J.R.

    1992-01-01

    This paper summarizes the findings of the Containment Working Group which met at the Workshop on Radioactive, Hazardous, and/or Mixed Waste Sludge Management. The Containment Working Group (CWG) examined the problems associated with providing adequate containment of waste forms from both short- and long-term storage. By its nature, containment encompasses a wide variety of waste forms, storage conditions, container types, containment schemes, and handling activities. A containment system can be anything from a 55-gal drum to a 100-ft-long underground vault. Because of the diverse nature of containment systems, the CWG chose to focus its limited time on broad issues that are applicable to the design of any containment system, rather than attempting to address problems specific to a particular containment system or waste-form type. Four major issues were identified by the CWG. They relate to: (1) service conditions and required system performance; (2) ultimate disposition; (3) cost and schedule; and (4) acceptance criteria, including quality assurance/quality control (QA/QC) concerns. All of the issues raised by the group are similar in that they all help to define containment system requirements

  13. Containment safety margins

    International Nuclear Information System (INIS)

    Von Riesemann, W.A.

    1980-01-01

    Objective of the Containment Safety Margins program is the development and verification of methodologies which are capable of reliably predicting the ultimate load-carrying capability of light water reactor containment structures under accident and severe environments. The program was initiated in June 1980 at Sandia and this paper addresses the first phase of the program which is essentially a planning effort. Brief comments are made about the second phase, which will involve testing of containment models

  14. FAUST/CONTAIN

    International Nuclear Information System (INIS)

    Cherdron, W.; Minges, J.; Sauter, H.; Schuetz, W.

    1995-01-01

    The FAUNA facility has been restructured after completion of the sodium fire experiments. It is now serving LWR research, cf. report II on program no. 32.21.02 concerning steam explosions. The CONTAIN code system for computing the thermodynamic, aerosol and radiological phenomena in a containment under severe accident conditions is being developed with a new to fission product release and transport. (orig.)

  15. Radiological containment handbook

    International Nuclear Information System (INIS)

    1982-10-01

    The purpose of this NUREG is to be used as a reference text. It is meant to be used by the working personnel as a guide for using temporary radiological containments. The installing group and health physics group may vary among organizations but responsibilities and duties will not change. It covers installation and inspection containments; working and operating guidelines; operating requirement; emergency procedures; and removal of containments

  16. Does atlas-based autosegmentation of neck levels require subsequent manual contour editing to avoid risk of severe target underdosage? A dosimetric analysis

    International Nuclear Information System (INIS)

    Voet, Peter W.J.; Dirkx, Maarten L.P.; Teguh, David N.; Hoogeman, Mischa S.; Levendag, Peter C.; Heijmen, Ben J.M.

    2011-01-01

    Background and purpose: To investigate the dosimetric impact of not editing auto-contours of the elective neck and organs at risk (OAR), generated with atlas-based autosegmentation (ABAS) (Elekta software) for head and neck cancer patients. Materials and methods: For nine patients ABAS auto-contours and auto-contours edited by two observers were available. Based on the non-edited auto-contours clinically acceptable IMRT plans were constructed (designated 'ABAS plans'). These plans were then evaluated for the two edited structure sets, by quantifying the percentage of the neck-PTV receiving more than 95% of the prescribed dose (V 95 ) and the near-minimum dose (D 99 ) in the neck PTV. Dice coefficients and mean contour distances were calculated to quantify the similarity of ABAS auto-contours with the structure sets edited by observer 1 and observer 2. To study the dosimetric importance of editing OAR auto-contours a new IMRT plan was generated for each patient-observer combination, based on the observer's edited CTV and the non-edited salivary gland auto-contours. For each plan mean doses for the non-edited glands were compared with doses for the same glands edited by the observer. Results: For both observers, edited neck CTVs were larger than ABAS auto-contours (p ≤ 0.04), by a mean of 8.7%. When evaluating ABAS plans on the PTVs of the edited structure sets, V 95 reduced by 7.2% ± 5.4% (1 SD) (p 99 was 14.2 Gy (range 1-54 Gy). Even for Dice coefficients >0.8 and mean contour distances 99 up to 11 Gy were observed. For treatment plans based on observer PTVs and non-edited auto-contoured salivary glands, the mean doses in the edited glands differed by only -0.6 Gy ± 1.0 Gy (p = 0.06). Conclusions: Editing of auto-contoured neck CTVs generated by ABAS is required to avoid large underdosages in target volumes. Often used similarity measures for evaluation of auto-contouring algorithms, such as dice coefficients, do not predict well for expected PTV underdose

  17. Reactor container

    International Nuclear Information System (INIS)

    Fukazawa, Masanori.

    1991-01-01

    A system for controlling combustible gases, it has been constituted at present such that the combustible gases are controlled by exhausting them to the wet well of a reactor container. In this system, however, there has been a problem, in a reactor container having plenums in addition to the wet well and the dry well, that the combustible gases in such plenums can not be controlled. In view of the above, in the present invention, suction ports or exhaust ports of the combustible gas control system are disposed to the wet well, the dry well and the plenums to control the combustible gases in the reactor container. Since this can control the combustible gases in the entire reactor container, the integrity of the reactor container can be ensured. (T.M.)

  18. Reactor containment

    International Nuclear Information System (INIS)

    Kawabe, Ryuhei; Yamaki, Rika.

    1990-01-01

    A water vessel is disposed and the gas phase portion of the water vessel is connected to a reactor container by a pipeline having a valve disposed at the midway thereof. A pipe in communication with external air is extended upwardly from the liquid phase portion to a considerable height so as to resist against the back pressure by a waterhead in the pipeline. Accordingly, when the pressure in the container is reduced to a negative level, air passes through the pipeline and uprises through the liquid phase portion in the water vessel in the form of bubbles and then flows into the reactor container. When the pressure inside of the reactor goes higher, since the liquid surface in the water vessel is forced down, water is pushed up into the pipeline. Since the waterhead pressure of a column of water in the pipeline and the pressure of the reactor container are well-balanced, gases in the reactor container are not leaked to the outside. Further, in a case if a great positive pressure is formed in the reactor container, the inner pressure overcomes the waterhead of the column of water, so that the gases containing radioactive aerosol uprise in the pipeline. Since water and the gases flow being in contact with each other, this can provide the effect of removing aerosol. (T.M.)

  19. ITER containment structures

    International Nuclear Information System (INIS)

    Sadakov, S.; Fauser, F.; Nelson, B.

    1991-01-01

    This document describes the results and recommendations of the Containment Structures Design Unit (CSDU) on the containment structures for ITER, made in the context of the Conceptual Design Phase. The document describes the following subsystems: (1) the primary vacuum vessel (VV), (2) the attaching locks (AL) of the invessel components, (3) the plasma passive and active stabilizers, (4) the cryostat vessel, and (5) the machine gravity supports. Although for most components reference designs were selected, for some of these alternative design options were described, because unresolved problems necessitate further research and development. Conclusions and future needs are summarized for each of the above subsystems: (1) a reference VV design was selected, while most critical VV future needs are the feasibility studies of manufacturing, assembly, and the repair/disassembly/reassembly by remote handling. Alternative, thin-wall options appear attractive and should be studied further during the Engineering Design Activities; (2) no reference design solution was selected for the AL system, as AL design requirements are extremely difficult and internally contradictory, while there is no existing tokamak precedent, but instead, five different approaches will be further researched early in the Engineering Design Phase; (3) significant progress is reported on passive loops, for which the ''twin-loops'' concept is ready to be advanced into the Engineering Design Phase, and on active coils, where a new coil positioning prevents interference with the blanket removal paths, and the current joints are located in a secondary vacuum or in the atmosphere of the reactor hall, repairable by remote handling; (4) a full metallic welded cryostat design with increased toroidal resistance was chosen, but with a design based on concrete with a thin inner metallic liner as a back-up in case detailed nuclear shielding requirements would force the cryostat to act as biological shield; (5) out

  20. Sharps container

    Science.gov (United States)

    Lee, Angelene M. (Inventor)

    1992-01-01

    This invention relates to a system for use in disposing of potentially hazardous items and more particularly a Sharps receptacle for used hypodermic needles and the like. A Sharps container is constructed from lightweight alodined nonmagnetic metal material with a cup member having an elongated tapered shape and length greater than its transverse dimensions. A magnet in the cup member provides for metal retention in the container. A nonmagnetic lid member has an opening and spring biased closure flap member. The flap member is constructed from stainless steel. A Velcro patch on the container permits selective attachment at desired locations.

  1. Containment Code Validation Matrix

    International Nuclear Information System (INIS)

    Chin, Yu-Shan; Mathew, P.M.; Glowa, Glenn; Dickson, Ray; Liang, Zhe; Leitch, Brian; Barber, Duncan; Vasic, Aleks; Bentaib, Ahmed; Journeau, Christophe; Malet, Jeanne; Studer, Etienne; Meynet, Nicolas; Piluso, Pascal; Gelain, Thomas; Michielsen, Nathalie; Peillon, Samuel; Porcheron, Emmanuel; Albiol, Thierry; Clement, Bernard; Sonnenkalb, Martin; Klein-Hessling, Walter; Arndt, Siegfried; Weber, Gunter; Yanez, Jorge; Kotchourko, Alexei; Kuznetsov, Mike; Sangiorgi, Marco; Fontanet, Joan; Herranz, Luis; Garcia De La Rua, Carmen; Santiago, Aleza Enciso; Andreani, Michele; Paladino, Domenico; Dreier, Joerg; Lee, Richard; Amri, Abdallah

    2014-01-01

    The Committee on the Safety of Nuclear Installations (CSNI) formed the CCVM (Containment Code Validation Matrix) task group in 2002. The objective of this group was to define a basic set of available experiments for code validation, covering the range of containment (ex-vessel) phenomena expected in the course of light and heavy water reactor design basis accidents and beyond design basis accidents/severe accidents. It was to consider phenomena relevant to pressurised heavy water reactor (PHWR), pressurised water reactor (PWR) and boiling water reactor (BWR) designs of Western origin as well as of Eastern European VVER types. This work would complement the two existing CSNI validation matrices for thermal hydraulic code validation (NEA/CSNI/R(1993)14) and In-vessel core degradation (NEA/CSNI/R(2001)21). The report initially provides a brief overview of the main features of a PWR, BWR, CANDU and VVER reactors. It also provides an overview of the ex-vessel corium retention (core catcher). It then provides a general overview of the accident progression for light water and heavy water reactors. The main focus is to capture most of the phenomena and safety systems employed in these reactor types and to highlight the differences. This CCVM contains a description of 127 phenomena, broken down into 6 categories: - Containment Thermal-hydraulics Phenomena; - Hydrogen Behaviour (Combustion, Mitigation and Generation) Phenomena; - Aerosol and Fission Product Behaviour Phenomena; - Iodine Chemistry Phenomena; - Core Melt Distribution and Behaviour in Containment Phenomena; - Systems Phenomena. A synopsis is provided for each phenomenon, including a description, references for further information, significance for DBA and SA/BDBA and a list of experiments that may be used for code validation. The report identified 213 experiments, broken down into the same six categories (as done for the phenomena). An experiment synopsis is provided for each test. Along with a test description

  2. Reactor container

    International Nuclear Information System (INIS)

    Kojima, Yoshihiro; Hosomi, Kenji; Otonari, Jun-ichiro.

    1997-01-01

    In the present invention, a catalyst for oxidizing hydrogen to be disposed in a reactor container upon rupture of pipelines of a reactor primary coolant system is prevented from deposition of water droplets formed from a reactor container spray to suppress elevation of hydrogen concentration in the reactor container. Namely, a catalytic combustion gas concentration control system comprises a catalyst for oxidizing hydrogen and a support thereof. In addition, there is also disposed a water droplet deposition-preventing means for preventing deposition of water droplets in a reactor pressure vessel on the catalyst. Then, the effect of the catalyst upon catalytic oxidation reaction of hydrogen can be kept high. The local elevation of hydrogen concentration can be prevented even upon occurrence of such a phenomenon that various kinds of mobile forces in the container such as dry well cooling system are lost. (I.S.)

  3. SEVERE ACCIDENT MANAGEMENT STATUS AT Loviisa

    International Nuclear Information System (INIS)

    Kymalainen, O.; Tuomisto, H.

    1997-01-01

    Some of the specific design features of IVO's Loviisa Plant, most notably the ice-condenser containment, strongly affect the plant response in a hypothetical core melt accident. They have together with the relatively stringent Finnish regulatory requirements forced IVO to develop a tailor made severe accident management strategy for Loviisa. The low design pressure of the ice-condenser containment complicates the design of the hydrogen management system. On the other hand, the ice-condensers and the water available from them are facilitating factors regarding in-vessel retention of corium by external cooling of reactor pressure vessel. This paper summarizes the Finnish severe accident requirements, IVO's approach to severe accidents, and its application to the Loviisa Plant

  4. CONTAIN independent peer review

    International Nuclear Information System (INIS)

    Boyack, B.E.; Corradini, M.L.; Khatib-Rahbar, M.; Loyalka, S.K.; Smith, P.N.

    1995-01-01

    The CONTAIN code was developed by Sandia National Laboratories under the sponsorship of the US Nuclear Regulatory Commission (NRC) to provide integrated analyses of containment phenomena. It is used to predict nuclear reactor containment loads, radiological source terms, and associated physical phenomena for a range of accident conditions encompassing both design-basis and severe accidents. The code's targeted applications include support for containment-related experimental programs, light water and advanced light water reactor plant analysis, and analytical support for resolution of specific technical issues such as direct containment heating. The NRC decided that a broad technical review of the code should be performed by technical experts to determine its overall technical adequacy. For this purpose, a six-member CONTAIN Peer Review Committee was organized and a peer review as conducted. While the review was in progress, the NRC issued a draft ''Revised Severe Accident Code Strategy'' that incorporated revised design objectives and targeted applications for the CONTAIN code. The committee continued its effort to develop findings relative to the original NRC statement of design objectives and targeted applications. However, the revised CONTAIN design objectives and targeted applications. However, the revised CONTAIN design objectives and targeted applications were considered by the Committee in assigning priorities to the Committee's recommendations. The Committee determined some improvements are warranted and provided recommendations in five code-related areas: (1) documentation, (2) user guidance, (3) modeling capability, (4) code assessment, and (5) technical assessment

  5. CONTAIN independent peer review

    Energy Technology Data Exchange (ETDEWEB)

    Boyack, B.E. [Los Alamos National Lab., NM (United States); Corradini, M.L. [Univ. of Wisconsin, Madison, WI (United States). Nuclear Engineering Dept.; Denning, R.S. [Battelle Memorial Inst., Columbus, OH (United States); Khatib-Rahbar, M. [Energy Research Inc., Rockville, MD (United States); Loyalka, S.K. [Univ. of Missouri, Columbia, MO (United States); Smith, P.N. [AEA Technology, Dorchester (United Kingdom). Winfrith Technology Center

    1995-01-01

    The CONTAIN code was developed by Sandia National Laboratories under the sponsorship of the US Nuclear Regulatory Commission (NRC) to provide integrated analyses of containment phenomena. It is used to predict nuclear reactor containment loads, radiological source terms, and associated physical phenomena for a range of accident conditions encompassing both design-basis and severe accidents. The code`s targeted applications include support for containment-related experimental programs, light water and advanced light water reactor plant analysis, and analytical support for resolution of specific technical issues such as direct containment heating. The NRC decided that a broad technical review of the code should be performed by technical experts to determine its overall technical adequacy. For this purpose, a six-member CONTAIN Peer Review Committee was organized and a peer review as conducted. While the review was in progress, the NRC issued a draft ``Revised Severe Accident Code Strategy`` that incorporated revised design objectives and targeted applications for the CONTAIN code. The committee continued its effort to develop findings relative to the original NRC statement of design objectives and targeted applications. However, the revised CONTAIN design objectives and targeted applications. However, the revised CONTAIN design objectives and targeted applications were considered by the Committee in assigning priorities to the Committee`s recommendations. The Committee determined some improvements are warranted and provided recommendations in five code-related areas: (1) documentation, (2) user guidance, (3) modeling capability, (4) code assessment, and (5) technical assessment.

  6. Reactor container cooling device

    Energy Technology Data Exchange (ETDEWEB)

    Ando, Koji; Kinoshita, Shoichiro

    1995-11-10

    The device of the present invention efficiently lowers pressure and temperature in a reactor container upon occurrence of a severe accident in a BWR-type reactor and can cool the inside of the container for a long period of time. That is, (1) pipelines on the side of an exhaustion tower of a filter portion in a filter bent device of the reactor container are in communication with pipelines on the side of a steam inlet of a static container cooling device by way of horizontal pipelines, (2) a back flow check valve is disposed to horizontal pipelines, (3) a steam discharge valve for a pressure vessel is disposed closer to the reactor container than the joint portion between the pipelines on the side of the steam inlet and the horizontal pipelines. Upon occurrence of a severe accident, when the pressure vessel should be ruptured and steams containing aerosol in the reactor core should be filled in the reactor container, the inlet valve of the static container cooling device is closed. Steams are flown into the filter bent device of the reactor container, where the aerosols can be removed. (I.S.).

  7. WASA-BOSS. Development and application of Severe Accident Codes. Evaluation and optimization of accident management measures. Subproject D. Study on water film cooling for PWR's passive containment cooling system. Final report

    International Nuclear Information System (INIS)

    Huang, Xi

    2016-07-01

    In the present study, a new phenomenological model was developed, to describe the water film flow under conditions of a passive containment cooling system (PCCS). The new model takes two different flow regimes into consideration, i.e. continuous water film and rivulets. For water film flow, the traditional Nusselt's was modified, to consider orientation angle and surface sheer stress. The transition from water film to rivulet as well as the structure of the stable rivulet at its onset point was modeled by using the minimum energy principle (MEP) combined with conservation equations. In addition, two different contact angles, i.e. advancing angle and retreating angle, were applied to take the hysteresis effect into consideration. The models of individual processes were validated as far as possible based on experimental data selected from open literature and from collaboration partner as well. With the models a new program module was developed and implemented into the COCOSYS program. The extended COCOSYS program was applied to analyze the containment behavior of the European generic containment and the performance of the passive containment cooling system ofthe AP1000. The results indicate clearly the importance of the new model and provide information for the optimization of the PCCS of AP1000.

  8. WASA-BOSS. Development and application of Severe Accident Codes. Evaluation and optimization of accident management measures. Subproject D. Study on water film cooling for PWR's passive containment cooling system. Final report

    Energy Technology Data Exchange (ETDEWEB)

    Huang, Xi

    2016-07-15

    In the present study, a new phenomenological model was developed, to describe the water film flow under conditions of a passive containment cooling system (PCCS). The new model takes two different flow regimes into consideration, i.e. continuous water film and rivulets. For water film flow, the traditional Nusselt's was modified, to consider orientation angle and surface sheer stress. The transition from water film to rivulet as well as the structure of the stable rivulet at its onset point was modeled by using the minimum energy principle (MEP) combined with conservation equations. In addition, two different contact angles, i.e. advancing angle and retreating angle, were applied to take the hysteresis effect into consideration. The models of individual processes were validated as far as possible based on experimental data selected from open literature and from collaboration partner as well. With the models a new program module was developed and implemented into the COCOSYS program. The extended COCOSYS program was applied to analyze the containment behavior of the European generic containment and the performance of the passive containment cooling system ofthe AP1000. The results indicate clearly the importance of the new model and provide information for the optimization of the PCCS of AP1000.

  9. Reactor container

    International Nuclear Information System (INIS)

    Furukawa, Hideyasu; Oyamada, Osamu; Uozumi, Hiroto.

    1976-01-01

    Purpose: To provide a container for a reactor provided with a pressure suppressing chamber pool which can prevent bubble vibrating load, particularly negative pressure generated at the time of starting to release exhaust from a main steam escape-safety valve from being transmitted to a lower liner plate of the container. Constitution: This arrangement is characterized in that a safety valve exhaust pool for main steam escape, in which a pressure suppressing chamber pool is separated and intercepted from pool water in the pressure suppressing chamber pool, a safety valve exhaust pipe is open into said safety valve exhaust pool, and an isolator member, which isolates the bottom liner plate in the pressure suppressing chamber pool from the pool water, is disposed on the bottom of the safety valve exhaust pool. (Nakamura, S.)

  10. CONTAIN calculations

    International Nuclear Information System (INIS)

    Scholtyssek, W.

    1995-01-01

    In the first phase of a benchmark comparison, the CONTAIN code was used to calculate an assumed EPR accident 'medium-sized leak in the cold leg', especially for the first two days after initiation of the accident. The results for global characteristics compare well with those of FIPLOC, MELCOR and WAVCO calculations, if the same materials data are used as input. However, significant differences show up for local quantities such as flows through leakages. (orig.)

  11. Reactor container

    International Nuclear Information System (INIS)

    Kato, Masami; Nishio, Masahide.

    1987-01-01

    Purpose: To prevent the rupture of the dry well even when the melted reactor core drops into a reactor pedestal cavity. Constitution: In a reactor container in which a dry well disposed above the reactor pedestal cavity for containing the reactor pressure vessel and a torus type suppression chamber for containing pressure suppression water are connected with each other, the pedestal cavity and the suppression chamber are disposed such that the flow level of the pedestal cavity is lower than the level of the pressure suppression water. Further, a pressure suppression water introduction pipeway for introducing the pressure suppression water into the reactor pedestal cavity is disposed by way of an ON-OFF valve. In case if the melted reactor core should fall into the pedestal cavity, the ON-OFF valve for the pressure suppression water introduction pipeway is opened to introduce the pressure suppression water in the suppression chamber into the pedestal cavity to cool the melted reactor core. (Ikeda, J.)

  12. Plasma container

    International Nuclear Information System (INIS)

    Ebisawa, Katsuyuki.

    1985-01-01

    Purpose: To enable to easily detect that the thickness of material to be abraded is reduced to an allowable limit from the outerside of the plasma container even during usual operation in a plasma vessel for a thermonuclear device. Constitution: A labelled material is disposed to the inside or rear face of constituent members of a plasma container undergoing the irradiation of plasma particles. A limiter plate to be abraded in the plasma container is composed of an armour member and heat removing plate, in which the armour member is made of graphite and heat-removing plate is made of copper. If the armour member is continuously abraded under the effect of sputtering due to plasma particles, silicon nitride embedded so far in the graphite at last appears on the surface of the limiter plate to undergo the impact shocks of the plasma particles. Accordingly, abrasion of the limiter material can be detected by a detector comprising gas chromatography and it can easily be detected from the outside of the plasma content even during normal operation. (Horiuchi, T.)

  13. Reactor container

    International Nuclear Information System (INIS)

    Shibata, Satoru; Kawashima, Hiroaki

    1984-01-01

    Purpose: To optimize the temperature distribution of the reactor container so as to moderate the thermal stress distribution on the reactor wall of LMFBR type reactor. Constitution: A good heat conductor (made of Al or Cu) is appended on the outer side of the reactor container wall from below the liquid level to the lower face of a deck plate. Further, heat insulators are disposed to the outside of the good heat conductor. Furthermore, a gas-cooling duct is circumferentially disposed at the contact portion between the good heat conductor and the deck plate around the reactor container. This enables to flow the cold heat from the liquid metal rapidly through the good heat conductor to the cooling duct and allows to maintain the temperature distribution on the reactor wall substantially linear even with the abrupt temperature change in the liquid metal. Further, by appending the good heat conductor covered with inactive metals not only on the outer side but also on the inside of the reactor wall to introduce the heat near the liquid level to the upper portion and escape the same to the cooling layer below the roof slab, the effect can be improved further. (Ikeda, J.)

  14. Fusion impulse containment

    International Nuclear Information System (INIS)

    Bohachevsky, I.O.

    1979-01-01

    The characteristics of impact fusion energy releases are not known sufficiently well to examine in detail specific containment vessel concepts or designs. Therefore it appears appropriate to formulate the impulse containment problem in general and to derive results in the form of explicit expressions from which magnitude estimates and parametric dependencies (trends) can be inferred conveniently and rapidly. In the following presentation we carry out this task using assumptions and approximations that are required to perform the analysis

  15. Nuclear reactor container

    International Nuclear Information System (INIS)

    Yamaki, Rika; Kawabe, Ryuhei.

    1989-01-01

    A venturi scrubber is connected to a nuclear reactor container. Gases containing radioactive aerosols in the container are introduced into the venturi scrubber in the form of a high speed stream under the pressure of the container. The radioactive aerosols are captured by inertia collision due to the velocity difference between the high speed gas stream and water droplets. In the case of the present invention, since the high pressure of the reactor container generated upon accident is utilized, compressor, etc. is no more required, thereby enabling to reduce the size of the aerosol removing device. Further, since no external power is used, the radioactive aerosols can be removed with no starting failure upon accidents. (T.M.)

  16. Westinghouse radiological containment guide

    International Nuclear Information System (INIS)

    Aitken, S.B.; Brown, R.L.; Cantrell, J.R.; Wilcox, D.P.

    1994-03-01

    This document provides uniform guidance for Westinghouse contractors on the implementation of radiological containments. This document reflects standard industry practices and is provided as a guide. The guidance presented herein is consistent with the requirements of the DOE Radiological Control Manual (DOE N 5480.6). This guidance should further serve to enable and encourage the use of containments for contamination control and to accomplish the following: Minimize personnel contamination; Prevent the spread of contamination; Minimize the required use of protective clothing and personal protective equipment; Minimize the generation of waste

  17. Westinghouse radiological containment guide

    Energy Technology Data Exchange (ETDEWEB)

    Aitken, S.B. [Idaho National Engineering Lab., Idaho Falls, ID (United States); Brown, R.L. [Westinghouse Hanford Co., Richland, WA (United States); Cantrell, J.R. [Westinghouse Savannah River Co., Aiken, SC (United States); Wilcox, D.P. [West Valley Nuclear Services Co., Inc., West Valley, NY (United States)

    1994-03-01

    This document provides uniform guidance for Westinghouse contractors on the implementation of radiological containments. This document reflects standard industry practices and is provided as a guide. The guidance presented herein is consistent with the requirements of the DOE Radiological Control Manual (DOE N 5480.6). This guidance should further serve to enable and encourage the use of containments for contamination control and to accomplish the following: Minimize personnel contamination; Prevent the spread of contamination; Minimize the required use of protective clothing and personal protective equipment; Minimize the generation of waste.

  18. CONTAIN calculations; CONTAIN-Rechnungen

    Energy Technology Data Exchange (ETDEWEB)

    Scholtyssek, W.

    1995-08-01

    In the first phase of a benchmark comparison, the CONTAIN code was used to calculate an assumed EPR accident `medium-sized leak in the cold leg`, especially for the first two days after initiation of the accident. The results for global characteristics compare well with those of FIPLOC, MELCOR and WAVCO calculations, if the same materials data are used as input. However, significant differences show up for local quantities such as flows through leakages. (orig.)

  19. Reactor container

    International Nuclear Information System (INIS)

    Hidaka, Masataka; Hatamiya, Shigeo; Kawasaki, Terufumi; Fukui, Toru; Suzuki, Hiroaki; Kataoka, Yoshiyuki; Kawabe, Ryuhei; Murase, Michio; Naito, Masanori.

    1990-01-01

    In order to suppress the pressure elevation in a reactor container due to high temperature and high pressure steams jetted out upon pipeway rupture accidents in the reactor container, the steams are introduced to a pressure suppression chamber for condensating them in stored coolants. However, the ability for suppressing the pressure elevation and steam coagulation are deteriorated due to the presence of inactive incondensible gases. Then, there are disposed a vent channel for introducing the steams in a dry well to a pressure suppression chamber in the reactor pressure vessel, a closed space disposed at the position lower than a usual liquid level, a first channel having an inlet in the pressure suppression chamber and an exit in the closed space and a second means connected by way of a backflow checking means for preventing the flow directing to the closed space. The first paths are present by plurality, a portion of which constitutes a syphon. The incondensible gases and the steams are discharged to the dry well at high pressure by using the difference of the water head for a long cooling time after the pipeway rupture accident. Then, safety can be improved without using dynamic equipments as driving source. (N.H.)

  20. Accumulation of immunoglobulin-containing cells in the gut mucosa and presence of faecal immunoglobulin in severe combined immunodeficient (scid) mice with T cell-induced inflammatory bowel disease (IBD)

    DEFF Research Database (Denmark)

    Bregenholt, S; Brimnes, J; Reimann, J

    1998-01-01

    and IgG2b were found to accumulate in colon segments displaying the most severe histopathology, including inflammatory cellular infiltration, epithelial hyperplasia and ulcerative lesions. Compared with colon segments of normal C.B-17 mice, the lesional scid colon shows increased levels of cells positive...

  1. ACE puts containment venting systems to the test

    International Nuclear Information System (INIS)

    Merilo, M.

    1990-01-01

    Filtered venting of reactor containments has received considerable attention recently as a method for avoiding containment failure due to overpressure during severe accidents. Several proposed filtration devices have been tested in the internationally sponsored Advanced Containment Experiments (ACE) programme, such that a self consistent comparison of the aerosol removal characteristics of these systems could be obtained. Considering the different design, requirements and operating conditions of the filter devices, a direct comparison is not possible, nor appropriate. Nevertheless, large scale models, using full scale elements of the various devices whenever feasible, have been tested with consistent mixtures of aerosols and carrier gases. (author)

  2. CANDU severe accident analysis

    International Nuclear Information System (INIS)

    Negut, Gheorghe; Catana, Alexandru; Prisecaru, Ilie; Dupleac, Daniel

    2007-01-01

    Romania is a EU member since January first 2007. This country faces now new challenges which imply also the nuclear power reactors now in operation. Romania operates since 1996 a CANDU nuclear power reactor and soon will start up a second unit. In EU PWR reactors are mostly operated, so that the Romania's reactors have to meet EU standards. Safety analysis guidelines require to model severe accidents for reactors of this type. Starting from previous studies a thermal-hydraulic model for a degraded CANDU core was developed. The initiating event is assumed to be a LOCA with simultaneous loss of moderator and coolant and the failure of emergency core cooling system (ECCS). This type of accident is likely to modify the reactor geometry and will lead to a severe accident development. When the coolant temperatures inside a pressure tube reaches 1000 deg. C, a contact between pressure tube and calandria tube occurs and the decay heat is transferred to the moderator. Due to the lack of cooling, the moderator eventually begins to boil and is expelled, through the calandria vessel relief ducts, into the containment. Therefore the calandria tubes (fuel channels) uncover, then disintegrate and fall down to the calandria vessel bottom. All the quantity of calandria moderator is vaporized and expelled, the debris will heat up and eventually boil. The heat accumulated in the molten debris will be transferred through the calandria vessel wall to the shield water tank surrounding the calandria vessel. The thermal hydraulics phenomena described above are modeled, analyzed and compared with the existing data. (authors)

  3. Two RNAs or DNAs May Artificially Fuse Together at a Short Homologous Sequence (SHS) during Reverse Transcription or Polymerase Chain Reactions, and Thus Reporting an SHS-Containing Chimeric RNA Requires Extra Caution

    Science.gov (United States)

    Xie, Bingkun; Yang, Wei; Ouyang, Yongchang; Chen, Lichan; Jiang, Hesheng; Liao, Yuying; Liao, D. Joshua

    2016-01-01

    Tens of thousands of chimeric RNAs have been reported. Most of them contain a short homologous sequence (SHS) at the joining site of the two partner genes but are not associated with a fusion gene. We hypothesize that many of these chimeras may be technical artifacts derived from SHS-caused mis-priming in reverse transcription (RT) or polymerase chain reactions (PCR). We cloned six chimeric complementary DNAs (cDNAs) formed by human mitochondrial (mt) 16S rRNA sequences at an SHS, which were similar to several expression sequence tags (ESTs).These chimeras, which could not be detected with cDNA protection assay, were likely formed because some regions of the 16S rRNA are reversely complementary to another region to form an SHS, which allows the downstream sequence to loop back and anneal at the SHS to prime the synthesis of its complementary strand, yielding a palindromic sequence that can form a hairpin-like structure.We identified a 16S rRNA that ended at the 4th nucleotide(nt) of the mt-tRNA-leu was dominant and thus should be the wild type. We also cloned a mouse Bcl2-Nek9 chimeric cDNA that contained a 5-nt unmatchable sequence between the two partners, contained two copies of the reverse primer in the same direction but did not contain the forward primer, making it unclear how this Bcl2-Nek9 was formed and amplified. Moreover, a cDNA was amplified because one primer has 4 nts matched to the template, suggesting that there may be many more artificial cDNAs than we have realized, because the nuclear and mt genomes have many more 4-nt than 5-nt or longer homologues. Altogether, the chimeric cDNAs we cloned are good examples suggesting that many cDNAs may be artifacts due to SHS-caused mis-priming and thus greater caution should be taken when new sequence is obtained from a technique involving DNA polymerization. PMID:27148738

  4. 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.)

  5. Containment vessel

    International Nuclear Information System (INIS)

    Zbirohowski-Koscia, K.F.; Roberts, A.C.

    1980-01-01

    A concrete containment vessel for nuclear reactors is disclosed that is spherical and that has prestressing tendons disposed in first, second and third sets, the tendons of each set being all substantially concentric and centred around a respective one of the three orthogonal axes of the sphere; the tendons of the first set being anchored at each end at a first anchor rib running around a circumference of the vessel, the tendons of the second set being anchored at each end at a second anchor rib running around a circumference of the sphere and disposed at 90 0 to the first rib, and the tendons of the third set being anchored some to the first rib and the remainder to the second rib. (author)

  6. Allergy in severe asthma

    NARCIS (Netherlands)

    Del Giacco, Stefano R.; Bakirtas, A.; Bel, E.; Custovic, A.; Diamant, Z.; Hamelmann, E.; Heffler, E.; Kalayci, O.; Saglani, S.; Sergejeva, S.; Seys, S.; Simpson, A.; Bjermer, Leif

    It is well recognized that atopic sensitization is an important risk factor for asthma, both in adults and in children. However, the role of allergy in severe asthma is still under debate. The term 'Severe Asthma' encompasses a highly heterogeneous group of patients who require treatment on steps

  7. Regulation Plans on Severe Accidents developed by KINS Severe Accident Regulation Preparation TFT

    International Nuclear Information System (INIS)

    Kim, Kyun Tae; Chung, Ku Young; Na, Han Bee

    2016-01-01

    Some nuclear power plants in Fukushima Daiichi site had lost their emergency reactor cooling function for long-time so the fuels inside the reactors were molten, and the integrity of containment was damaged. Therefore, large amount of radioactive material was released to environment. Because the social and economic effects of severe accidents are enormous, Korean Government already issued 'Severe Accident Policy' in 2001 which requires nuclear power plant operators to set up 'Quantitative Safety Goal', to do 'Probabilistic Safety Analysis', to install 'Severe Accident Countermeasures' and to make 'Severe Accident Management Plan'. After the Fukushima disaster, a Special Safety Inspection was performed for all operating nuclear power plants of Korea. The inspection team from industry, academia, and research institutes assessed Korean NPPs capabilities to cope with or respond to severe accidents and emergency situation caused by natural disasters such as a large earthquake or tsunami. As a result of the special inspection, about 50 action items were identified to increase the capability to cope with natural disaster and severe accidents. Nuclear Safety Act has been amended to require NPP operators to submit Accident Management Plant as part of operating license application. The KINS Severe Accident Regulation Preparation TFT had first investigated oversea severe accident regulation trend before and after the Fukushima accident. Then, the TFT has developed regulation draft for severe accidents such as Severe accident Management Plans, the required design features for new NPPs to prevent severe accident against multiple failures and beyond-design external events, countermeasures to mitigate severe accident and to keep the integrity of containment, and assessment methodology on safety assessment plan and probabilistic safety assessment

  8. Regulation Plans on Severe Accidents developed by KINS Severe Accident Regulation Preparation TFT

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Kyun Tae; Chung, Ku Young; Na, Han Bee [KINS, Daejeon (Korea, Republic of)

    2016-05-15

    Some nuclear power plants in Fukushima Daiichi site had lost their emergency reactor cooling function for long-time so the fuels inside the reactors were molten, and the integrity of containment was damaged. Therefore, large amount of radioactive material was released to environment. Because the social and economic effects of severe accidents are enormous, Korean Government already issued 'Severe Accident Policy' in 2001 which requires nuclear power plant operators to set up 'Quantitative Safety Goal', to do 'Probabilistic Safety Analysis', to install 'Severe Accident Countermeasures' and to make 'Severe Accident Management Plan'. After the Fukushima disaster, a Special Safety Inspection was performed for all operating nuclear power plants of Korea. The inspection team from industry, academia, and research institutes assessed Korean NPPs capabilities to cope with or respond to severe accidents and emergency situation caused by natural disasters such as a large earthquake or tsunami. As a result of the special inspection, about 50 action items were identified to increase the capability to cope with natural disaster and severe accidents. Nuclear Safety Act has been amended to require NPP operators to submit Accident Management Plant as part of operating license application. The KINS Severe Accident Regulation Preparation TFT had first investigated oversea severe accident regulation trend before and after the Fukushima accident. Then, the TFT has developed regulation draft for severe accidents such as Severe accident Management Plans, the required design features for new NPPs to prevent severe accident against multiple failures and beyond-design external events, countermeasures to mitigate severe accident and to keep the integrity of containment, and assessment methodology on safety assessment plan and probabilistic safety assessment.

  9. System 80+ design features for severe accident prevention and mitigation

    International Nuclear Information System (INIS)

    Jacob, M.C.; Schneider, R.E.; Finnicum, D.J.

    1993-01-01

    ABB-CE, in cooperation with the US Department of Energy, is working to develop and certify the System 80+ design, which is ABB-CE's standardized evolutionary Advanced Light Water Reactor (ALWR) design. It incorporates design enhancements based on Probabilistic Risk Assessment (PRA) insights, guidance from the EPRI's Utility Requirements Document, and US NRC's Severe Accident Policy. Major severe accident prevention and mitigation design features of the system is discussed along with its conformance to EPRI URD guidance, as applicable. Computer simulation of a best estimate severe accident scenario is presented to illustrate the acceptable containment performance of the design. It is concluded that by considering severe accident prevention and mitigation early in the design process, the System 80+ design represents a robust plant design that has low core damage frequencies, low containment conditional failure probabilities, and acceptable deterministic containment performance under severe accident conditions

  10. Development on design methodology of PWR passive containment system

    International Nuclear Information System (INIS)

    Lee, Seong Wook

    1998-02-01

    The containment is the most important barrier against the release of radioactive materials into the environment during accident conditions of nuclear power plants. Therefore the development of a reliable containment cooling system is one of key areas in advanced reactor development. To enhance the safety of the containment system, many new containment system designs have been proposed and developed in the world. Several passive containment cooling system (PCCS) concepts for both steel and concrete containment systems are overviewed and assessed comparatively. Major concepts considered are: (a) the spray of water on the outer surface of a steel containment from an elevated tank, (b) an external moat for a steel containment, (c) a suppression pool for a concrete containment, and (d) combination of the internal spray and internal or external condensers for a concrete containment. Emphasis is given to the heat removal principles, the required heat transfer area, system complexity and operational reliability. As one of conceptual design steps of containment, a methodology based on scaling principles is proposed to determine the containment size according to the power level. The AP600 containment system is selected as the reference containment to which the scaling laws are applied. Governing equations of containment pressure are set up in consideration of containment behavior in accident conditions. Then, the dimensionless numbers, which characterize the containment phenomena, are derived for the blowdown dominant and decay heat dominant stage, respectively. The important phenomena in blowdown stage are mass and energy sources and their absorption in containment atmosphere or containment structure, while heat transfer to the outer environment becomes important in decay heat stage. Based on their similarity between the prototype and the model, the containment sizes are determined for higher power levels and are compared with the SPWR containment design values available

  11. Containment concepts assessment for the SEAFP reactor

    International Nuclear Information System (INIS)

    Di Pace, L.; Natalizio, A.

    2000-01-01

    A simple methodology has been developed for making relative comparisons of potential containment designs for future fusion reactors. The assessment methodology requires only conceptual design information. The application of this methodology, at the early stages of a fusion reactor design, provides designers useful information regarding the suitability of various containment designs and design features. Because the radiation hazard from the operation of future fusion power reactors is expected to be low, the containment design, in addition to public safety, needs to take into account worker safety considerations, as well as factors important to the reliable and economical operation of the power plant. Several containment concepts have been assessed with a methodology that takes into account public safety, worker safety, operability and maintainability as well as cost. This paper describes this methodology and presents the results of the assessment. The paper concludes that, to obtain a containment design that is optimised with respect to safety, operational and cost factors, designers should focus on a containment that is conceptually simple-that is, one utilising a single, large containment building without relying on special features such as expansion volumes, pressure suppression pools or spray systems

  12. Component nuclear containment structure

    International Nuclear Information System (INIS)

    Harstead, G.A.

    1979-01-01

    The invention described is intended for use primarily as a nuclear containment structure. Such structures are required to surround the nuclear steam supply system and to contain the effects of breaks in the nuclear steam supply system, or i.e. loss of coolant accidents. Nuclear containment structures are required to withstand internal pressure and temperatures which result from loss of coolant accidents, and to provide for radiation shielding during operation and during the loss of coolant accident, as well as to resist all other applied loads, such as earthquakes. The nuclear containment structure described herein is a composite nuclear containment structure, and is one which structurally combines two previous systems; namely, a steel vessel, and a lined concrete structure. The steel vessel provides strength to resist internal pressure and accommodate temperature increases, the lined concrete structure provides resistance to internal pressure by having a liner which will prevent leakage, and which is in contact with the concrete structure which provides the strength to resist the pressure

  13. Kinetic studies of the radical oxidation in gaseous phase of organic iodides and of the formation of iodine oxide particles under the simulated conditions of a nuclear reactor containment submitted to a severe accident

    International Nuclear Information System (INIS)

    Zhang, S.

    2012-01-01

    Within the framework of the research in the nuclear reactor safety field, the iodine oxides formation by organic iodides destruction in the containment has been studied with the means of the atmospheric chemistry field. The destruction kinetics and their activation energy of organic iodides by . OH and . O radical has been quantified by a Flash Photolysis system able to monitor the oxidant radicals by resonance fluorescence. Those results have been published and some of them for the first time in the literature. The mechanisms leading to the organic iodides destruction are either by a hydrogen atom abstraction, either by the formation of a complex, depending on the organic iodide involved. Then, certain kinetics reactions have been updated in the IODAIR code. Other reactions have been added based on the recent literature available. A comparison of the kinetics destruction of CH 3 I by . OH and . O with IODAIR and the global kinetics of destruction in ASTEC/IODE showed a difference of about 2 which shows the importance of these two radicals (and mainly . O) in those destruction processes. The other main path of destruction would be by electron radiation. Other radicals like . H and . N would not contribute significantly to organic iodides destruction. A sensitivity analysis highlighted that organic iodides would mostly be destroyed into iodine oxides with a almost complete conversion within a few hours. Finally, an atmospheric chamber has been used to quantify iodine oxides growth, density and composition. Under the conditions studied, their formation is fast. Particles sizes of about 200-400 nm are formed within a few hours. The main parameters influencing their growth are the relative humidity and the presence of ozone (whose function is to create . O and . OH radicals). (author)

  14. Bellefonte primary containment structure

    International Nuclear Information System (INIS)

    Olyniec, J.H.

    1981-01-01

    Construction of the reactor building primary containment structure at the Bellefonte Nuclear Plant involved several specialized construction techniques. This two unit plant is one of the nine nuclear units at six different sites now under construction by the Tennessee Valley Authority (TVA). The post-Tensioned, cast-in-place interior steel lined containment structure is unique within TVA. Problems during construction were identified at weekly planning meetings, and options were discussed. Close coordination between craft supervisors and on-site engineering personnel drew together ''hands-on''experience and technical background. Details of the construction techniques, problems, and solutions are presented

  15. Effect of a dietary supplement containing kurozu (a Japanese traditional health drink concentrate on several obesity-related parameters in obese Japanese adults: a randomized, double-blind, placebo-controlled trial

    Directory of Open Access Journals (Sweden)

    Naobumi Hamadate

    2013-08-01

    Full Text Available ABSTRACTObjective: This study was undertaken to examine the Kurozu concentrate (KC based dietary supplement on several obesity-related parameters in obese Japanese male and female adults.Background: Kurozu, which is a specific type of rice vinegar produced by fermentation of unpolished brown rice, has long been used as a traditional health food and folk medicine in Japan. A recent animal study and our preliminary human study suggest that the KC supplement has potential for use in the management of obesity.Materials and Methods: A 12-week, randomized, double-blind, placebo-controlled trial was conducted involving 48 Japanese adult subjects (28 males and 20 females with obesity. Subjects were either assigned to the group consuming the KC supplement for 12 weeks (870mg/day; 480 mg/day as KC (n=24; 14 males and 10 females or the placebo group (n=24; 14 males and 10 females. All test participants were assessed using several obesity-related parameters, including body weight, BMI, waist circumference, and abdominal fat computed tomography (CT sections. These measurements took place at baseline and at week 12. Results: At week 12, a significant decrease in body weight (P<0.043 and nearly significant decreased values of BMI (P=0.052 were observed in the KC group compared to the placebo group. The reduction in waist circumference at week 12 within the KC group was not significantly greater than the placebo group. Examination of abdominal CT sections around the navel indicated that, although most of the values of the total fat area, subcutaneous fat area, and visceral fat area for both of the placebo and KC groups significantly increased during the 12-week intervention, the magnitude of increase in the total fat area for all subjects and that of the total fat area, subcutaneous fat area, and visceral fat area for females on one or more of three CT sections were significantly lower in the KC group than the placebo group (P<0.05.Conclusion: Although the

  16. Translational and structural requirements of the early nodulin gene enod40, a short-open reading frame-containing RNA, for elicitation of a cell-specific growth response in the alfalfa root cortex.

    Science.gov (United States)

    Sousa, C; Johansson, C; Charon, C; Manyani, H; Sautter, C; Kondorosi, A; Crespi, M

    2001-01-01

    A diversity of mRNAs containing only short open reading frames (sORF-RNAs; encoding less than 30 amino acids) have been shown to be induced in growth and differentiation processes. The early nodulin gene enod40, coding for a 0.7-kb sORF-RNA, is expressed in the nodule primordium developing in the root cortex of leguminous plants after infection by symbiotic bacteria. Ballistic microtargeting of this gene into Medicago roots induced division of cortical cells. Translation of two sORFs (I and II, 13 and 27 amino acids, respectively) present in the conserved 5' and 3' regions of enod40 was required for this biological activity. These sORFs may be translated in roots via a reinitiation mechanism. In vitro translation products starting from the ATG of sORF I were detectable by mutating enod40 to yield peptides larger than 38 amino acids. Deletion of a Medicago truncatula enod40 region between the sORFs, spanning a predicted RNA structure, did not affect their translation but resulted in significantly decreased biological activity. Our data reveal a complex regulation of enod40 action, pointing to a role of sORF-encoded peptides and structured RNA signals in developmental processes involving sORF-RNAs.

  17. Material containment enclosure

    International Nuclear Information System (INIS)

    Carlson, D.O.

    1993-01-01

    An isolation enclosure and a group of isolation enclosures are described which are useful when a relatively large containment area is required. The enclosure is in the form of a ring having a section removed so that a technician may enter the center area of the ring. In a preferred embodiment, an access zone is located in the transparent wall of the enclosure and extends around the inner perimeter of the ring so that a technician can insert his hands into the enclosure to reach any point within. The inventive enclosures provide more containment area per unit area of floor space than conventional material isolation enclosures. 3 figures

  18. Containment long-term operational integrity--a 1988 status report

    International Nuclear Information System (INIS)

    Sammataro, R.F.

    1988-01-01

    Design and in-service codes and standards provide a comprehensive set of requirements for containment design, construction, inspection, testing and repair. Metal and concrete containments must be designed, fabricated, constructed, inspected, tested and maintained to quality standards commensurate with the importance of the safety function to be performed. Periodic integrated leak rate tests are required to assure that containments continue to meet allowable leakage limits. Although overall performance has been quite good to date, several major containment aging and degradation mechanisms have been identified. Two pilot plant life extension studies, one for a boiling water reactor and one for a pressurized water reactor, serve as models for extending the operational integrity of present containments in the United States. Research and testing programs for determining the ultimate pressure capacity and failure mechanisms for containments under severe loading conditions and studied for extending the life of current plants beyond the present 40 year licensed lifetime are underway. This paper presents an overview of containment designs in the USA and a discussion of the regulatory and ASME Code requirements for the design, construction, in-service inspection, testing and repair for containments. Findings for containments from the pilot plant life extension studies and the ongoing containment research and testing programs are also discussed. The regulatory and ASME Code requirements for design, construction, in-service inspection and periodic integrated leakage testing together with recommendations from the plant life extension studies and containment integrity research and testing provide a basis for continued containment long-term operational integrity

  19. 46 CFR 160.023-6 - Container.

    Science.gov (United States)

    2010-10-01

    ... 46 Shipping 6 2010-10-01 2010-10-01 false Container. 160.023-6 Section 160.023-6 Shipping COAST... Container. (a) General. The container for storing the signals on lifeboats and liferafts is not required to be of a special design or be approved by the Coast Guard. The container must meet the requirements in...

  20. Steel containment buckling

    International Nuclear Information System (INIS)

    Butler, T.A.; Baker, W.E.

    1987-01-01

    Two aspects of buckling of a free-standing nuclear containment building were investigated in a combined experimental and analytical program. In the first part of the study, the response of a scale model of a containment building to dynamic base excitation is investigated. A simple harmonic signal was used for preliminary studies followed by experiments with scaled earthquake signals as the excitation source. The experiments and accompanying analyses indicate that the scale model response to earthquake-type excitations is very complex and that current analytical methods may require that a dynamic capacity reduction factor be incorporated. The second part of the study quantified the effects of framing at large penetrations on the static buckling capacity of scale model containments. Results show little effect from the framing for the scale models constructed from the polycarbonate, Lexan. However, additional studies with a model constructed of the prototypic steel material are recommended. (orig.)

  1. Steel containment buckling

    International Nuclear Information System (INIS)

    Butler, T.A.; Baker, W.E.

    1986-01-01

    Two aspects of buckling of a free-standing nuclear steel containment building were investigated in a combined experimental and analytical program. In the first part of the study, the response of a scale model of a containment building to dynamic base excitation is investigated. A simple harmonic signal was used for preliminary studies followed by experiments with scaled earthquake signals as the excitation source. The experiments and accompanying analyses indicate that the scale model response to earthquake-type excitations is very complex and that current analytical methods may require a dynamic capacity reduction factor to be incorporated. The second part of the study quantified the effects of framing at large penetrations on the static buckling capacity of scale model containments. Results show little effect from the framing for the scale models constructed from the polycarbonate, Lexan. However, additional studies with a model constructed of the prototypic steel material are suggested

  2. Materials designed for containments

    International Nuclear Information System (INIS)

    Piehl, K.H.

    1976-01-01

    The present article points out that high-tensile fine-grained steels have been used successfully in the construction of reactor containments, spherical gasometers, and pressure vessels. It has been confirmed that their use requires safety measures concerning lay out and production. Viscosity properties of high-tensile, fine-grained steels can be improved significantly by means of electroslag remelting. The extent to which this improvement influences the heat-affected zone is being examined. (orig./RW) [de

  3. 75 FR 33705 - Pesticide Management and Disposal; Standards for Pesticide Containers and Containment; Change to...

    Science.gov (United States)

    2010-06-15

    ... Pesticide Management and Disposal; Standards for Pesticide Containers and Containment; Change to Labeling... the pesticide container and containment regulations to provide a 4-month extension of the 40 CFR 156... pesticide labels to comply with the label requirements in the container and containment regulations. DATES...

  4. 75 FR 26268 - Agency Information Collection Activities: Permit To Transfer Containers to a Container Station

    Science.gov (United States)

    2010-05-11

    ... Activities: Permit To Transfer Containers to a Container Station AGENCY: U.S. Customs and Border Protection... information collection requirement concerning the: Permit to Transfer Containers to a Container Station. This... information collection: Title: Permit to Transfer Containers to a Container Station. OMB Number: 1651-0049...

  5. Design and implementation of check out station for large container inspection system

    International Nuclear Information System (INIS)

    Yao Dongsheng; Gao Wenhuan; Kang Kejun

    1997-01-01

    In Large Container Inspection System (LCIS), Check Out Station (COS) is in charge of deciding whether a container is allowed to pass or has to be opened for checking. Several different top level architecture designs for COS are discussed and analyzed according to the practical requirements of COS

  6. Residual Stresses In 3013 Containers

    International Nuclear Information System (INIS)

    Mickalonis, J.; Dunn, K.

    2009-01-01

    The DOE Complex is packaging plutonium-bearing materials for storage and eventual disposition or disposal. The materials are handled according to the DOE-STD-3013 which outlines general requirements for stabilization, packaging and long-term storage. The storage vessels for the plutonium-bearing materials are termed 3013 containers. Stress corrosion cracking has been identified as a potential container degradation mode and this work determined that the residual stresses in the containers are sufficient to support such cracking. Sections of the 3013 outer, inner, and convenience containers, in both the as-fabricated condition and the closure welded condition, were evaluated per ASTM standard G-36. The standard requires exposure to a boiling magnesium chloride solution, which is an aggressive testing solution. Tests in a less aggressive 40% calcium chloride solution were also conducted. These tests were used to reveal the relative stress corrosion cracking susceptibility of the as fabricated 3013 containers. Significant cracking was observed in all containers in areas near welds and transitions in the container diameter. Stress corrosion cracks developed in both the lid and the body of gas tungsten arc welded and laser closure welded containers. The development of stress corrosion cracks in the as-fabricated and in the closure welded container samples demonstrates that the residual stresses in the 3013 containers are sufficient to support stress corrosion cracking if the environmental conditions inside the containers do not preclude the cracking process.

  7. Ultimate pressure capacity of CANDU 6 containment structures

    International Nuclear Information System (INIS)

    Radulescu, J.P.; Pradolin, L.; Mamet, J.C.

    1997-01-01

    This paper summarizes the analytical work carried out and the results obtained when determining the ultimate pressure capacity (UPC) of the containment structures of CANDU 6 nuclear power plants. The purpose of the analysis work was to demonstrate that such containment structures are capable of meeting design requirements under the most severe accident conditions. For this concrete vessel subjected to internal pressure, the UPC was defined as the pressure causing through cracking in the concrete. The present paper deals with the overall behaviour of the containment. The presence of openings, penetrations and the ultimate pressure of the airlocks were considered separately. (author)

  8. Severe accident management program at Cofrentes Nuclear Power Plant

    International Nuclear Information System (INIS)

    Borondo, L.; Serrano, C.; Fiol, M.J.; Sanchez, A.

    2000-01-01

    Cofrentes Nuclear Power Plant (GE BWR/6) has implemented its specific Severe Accident Management Program within this year 2000. New organization and guides have been developed to successfully undertake the management of a severe accident. In particular, the Technical Support Center will count on a new ''Severe Accident Management Team'' (SAMT) which will be in charge of the Severe Accident Guides (SAG) when Control Room Crew reaches the Emergency Operation Procedures (EOP) step that requires containment flooding. Specific tools and training have also been developed to help the SAMT to mitigate the accident. (author)

  9. Development and analysis of vent-filtered containment conceptual designs

    International Nuclear Information System (INIS)

    Benjamin, A.S.; Walling, H.C.

    1980-01-01

    Conceptual filtered-vented containment systems have been postulated for a reference large, dry, pressurized water reactor containment, and the systems have been analyzed to determine design parameters, actuation/operation requirements, and overall feasibility. The primary design challenge has been found to emanate from pressure spikes caused by core debris bed interactions with water and by hydrogen deflagrations. Circumvention of the pressure spikes may require a more complicated actuation logic than has previously been considered. Otherwise, major reductions in consequences for certain severe accidents appear to be possible with relatively simple systems. A probabilistic assessment of competing risks remains to be performed

  10. [Severe rhabdomyolysis secondary to severe hypernatraemic dehydration].

    Science.gov (United States)

    Mastro-Martínez, Ignacio; Montes-Arjona, Ana María; Escudero-Lirio, Margarita; Hernández-García, Bárbara; Fernández-Cantalejo Padial, José

    2015-01-01

    Rhabdomyolysis is a rare paediatric condition. The case is presented of a patient in whom this developed secondary to severe hypernatraemic dehydration following acute diarrhoea. Infant 11 months of age who presented with vomiting, fever, diarrhoea and anuria for 15 hours. Parents reported adequate preparation of artificial formula and oral rehydration solution. He was admitted with malaise, severe dehydration signs and symptoms, cyanosis, and low reactivity. The laboratory tests highlighted severe metabolic acidosis, hypernatraemia and pre-renal kidney failure (Sodium [Na] plasma 181 mEq/L, urine density> 1030). He was managed in Intensive Care Unit with gradual clinical and renal function improvement. On the third day, slight axial hypotonia and elevated cell lysis enzymes (creatine phosphokinase 75,076 IU/L) were observed, interpreted as rhabdomyolysis. He was treated with intravenous rehydration up to 1.5 times the basal requirements, and he showed a good clinical and biochemical response, being discharged 12 days after admission without motor sequelae. Severe hypernatraemia is described as a rare cause of rhabdomyolysis and renal failure. In critically ill patients, it is important to have a high index of suspicion for rhabdomyolysis and performing serial determinations of creatine phosphokinase for early detection and treatment. Copyright © 2015 Sociedad Chilena de Pediatría. Publicado por Elsevier España, S.L.U. All rights reserved.

  11. Special closures for steel drum shipping containers

    International Nuclear Information System (INIS)

    Bonzon, L.L.; Otts, J.V.

    1976-01-01

    The objective of this program was to develop special lid closures for typical, steel drum, radioactive material shipping containers. Previous experience and testing had shown that the existing container was adequate to withstand the required environmental tests for certification, but that the lid and closure were just marginally effective. Specifically, the lid closure failed to consistently maintain a tight seal between the container and the lid after drop tests, thus causing the package contents to be vulnerable in the subsequent fire test. Recognizing the deficiency, the United States Energy Research and Development Administration requested the development of new closure(s) which would: (1) be as strong and resistant to a drop as the bottom of the container; (2) have minimal economic impact on the overall container cost; (3) maximize the use of existing container designs; (4) consider crush loads; and (5) result in less dependence on personnel and loading procedures. Several techniques were evaluated and found to be more effective than the standard closure mechanism. Of these, three new closure techniques were designed, fabricated, and proven to be structurally adequate to provide containment when a 454-kg drum was drop tested from 9.14-m onto an unyielding surface. The three designs were: (1) a 152-mm long lid extension or skirt welded to the standard drum lid, (2) a separate inner lid, with 152-mm long skirt and (3) C-clamps used at the container-lid interface. Based upon structural integrity, economic impact, and minimal design change, the lid extension is the recommended special closure

  12. Life extension of containment structures of Indian PHWRs

    International Nuclear Information System (INIS)

    Roy, Raghupati; Garg, R.P.; Verma, U.S.P.

    2006-01-01

    Containment structures prevent radioactivity release in the event of any postulated Design Basis Accident (DBA) so that the level of radiation in the external environment is within acceptable limits. Containment structures of Indian PHWRs are typically unlined prestressed concrete structures, which are required to maintain its leak tightness characteristics and strength under DBA during the life of the structure. As nuclear power plant structures age, a number of degradation mechanisms begin to affect critical containment structure. Depending on the type and severity of these degradation mechanisms, its adverse effect on the leak tightness and pressure carrying capacity can be significant. Since the containment structures of Indian PHWRs are unlined, the leak tightness characteristics are solely dependent on the concrete properties and the prestressing material. Prestressing, which is introduced to control the deformation and strength requirement, is affected due to aging. Hence, adequacy of prestressing during the life of the structure to withstand internal pressure and the related leak tightness must be ensured for life extension of prestressed concrete containment structure in view of their significant long term losses. Prevention of corrosion in prestressing steel and assessment of the same at the end of extended design life of the structure, require utmost attention in view of their catastrophic nature of failure. This paper describes the various degradation mechanisms pertaining to concrete and their effect on the leak tightness characteristics and the strength requirement. The issues related to prestressing are also discussed in detail in this paper. The requirement of periodic monitoring of the containment structure for assessing its deformation and leak tightness characteristics and development of database for life extension of containment structure is also addressed in this paper. This paper also discusses the various provisions and measures, which are

  13. ACR-1000 design provisions for severe accidents

    International Nuclear Information System (INIS)

    Popov, N.K.; Santamaura, P.; Shapiro, H.; Snell, V.G.

    2006-01-01

    Atomic Energy of Canada Limited (AECL) developed the Advanced CANDU Reactor-700 (ACR-700) as an evolutionary advancement of the current CANDU 6 reactor. As a further advancement of the ACR design, AECL is currently developing the ACR-1000 for the Canadian and international market. The ACR-1000 is aimed at producing electrical power for a capital cost and a unit-energy cost significantly less than that of the current generation of operating nuclear plants, while achieving enhanced safety features, shorter construction schedule, high plant capacity factor, improved operations and maintenance, and increased operating life. The reference ACR-1000 plant design is based on an integrated two-unit plant, using enriched fuel and light-water coolant, with each unit having a nominal gross output of about 1200 MWe. The ACR-1000 design meets Canadian regulatory requirements and follows established international practice with respect to severe accident prevention and mitigation. This paper presents the ACR-1000 features that are designed to mitigate limited core damage and severe core damage states, including core retention within vessel, core damage termination, and containment integrity maintenance. While maintaining existing structures of CANDU reactors that provide inherent prevention and retention of core debris, the ACR-1000 design includes additional features for prevention and mitigation of severe accidents. Core retention within vessel in CANDU-type reactors includes both retention within fuel channels, and retention within the calandria vessel. The ACR-1000 calandria vessel design permits for passive rejection of decay heat from the moderator to the shield water. Also, the calandria vessel is designed for debris retention by minimizing penetrations at the bottom periphery and by accommodating thermal and weight loads of the core debris. The ACR-1000 containment is required to withstand external events such as earthquakes, tornados, floods and aircraft crashes

  14. Direct containment heating models in the CONTAIN code

    International Nuclear Information System (INIS)

    Washington, K.E.; Williams, D.C.

    1995-08-01

    The potential exists in a nuclear reactor core melt severe accident for molten core debris to be dispersed under high pressure into the containment building. If this occurs, the set of phenomena that result in the transfer of energy to the containment atmosphere and its surroundings is referred to as direct containment heating (DCH). Because of the potential for DCH to lead to early containment failure, the U.S. Nuclear Regulatory Commission (USNRC) has sponsored an extensive research program consisting of experimental, analytical, and risk integration components. An important element of the analytical research has been the development and assessment of direct containment heating models in the CONTAIN code. This report documents the DCH models in the CONTAIN code. DCH models in CONTAIN for representing debris transport, trapping, chemical reactions, and heat transfer from debris to the containment atmosphere and surroundings are described. The descriptions include the governing equations and input instructions in CONTAIN unique to performing DCH calculations. Modifications made to the combustion models in CONTAIN for representing the combustion of DCH-produced and pre-existing hydrogen under DCH conditions are also described. Input table options for representing the discharge of debris from the RPV and the entrainment phase of the DCH process are also described. A sample calculation is presented to demonstrate the functionality of the models. The results show that reasonable behavior is obtained when the models are used to predict the sixth Zion geometry integral effects test at 1/10th scale

  15. Direct containment heating models in the CONTAIN code

    Energy Technology Data Exchange (ETDEWEB)

    Washington, K.E.; Williams, D.C.

    1995-08-01

    The potential exists in a nuclear reactor core melt severe accident for molten core debris to be dispersed under high pressure into the containment building. If this occurs, the set of phenomena that result in the transfer of energy to the containment atmosphere and its surroundings is referred to as direct containment heating (DCH). Because of the potential for DCH to lead to early containment failure, the U.S. Nuclear Regulatory Commission (USNRC) has sponsored an extensive research program consisting of experimental, analytical, and risk integration components. An important element of the analytical research has been the development and assessment of direct containment heating models in the CONTAIN code. This report documents the DCH models in the CONTAIN code. DCH models in CONTAIN for representing debris transport, trapping, chemical reactions, and heat transfer from debris to the containment atmosphere and surroundings are described. The descriptions include the governing equations and input instructions in CONTAIN unique to performing DCH calculations. Modifications made to the combustion models in CONTAIN for representing the combustion of DCH-produced and pre-existing hydrogen under DCH conditions are also described. Input table options for representing the discharge of debris from the RPV and the entrainment phase of the DCH process are also described. A sample calculation is presented to demonstrate the functionality of the models. The results show that reasonable behavior is obtained when the models are used to predict the sixth Zion geometry integral effects test at 1/10th scale.

  16. Reactor container facility

    International Nuclear Information System (INIS)

    Saito, Takashi; Nagasaka, Hideo.

    1990-01-01

    A dry-well pool for spontaneously circulating stored pool water and a suppression pool for flooding a pressure vessel by feeding water, when required, to a flooding gap by means of spontaneous falling upto the flooding position, thereby flooding the pressure vessel are contained at the inside of a reactor container. Thus, when loss of coolant accidents such as caused by main pipe rupture accidents should happen, pool water in the suppression pool is supplied to the flooding gap by spontaneously falling. Further, if the flooding water uprises exceeding a predetermined level, the flooding gap is in communication with the dry-well pool at the upper and the lower portions respectively. Accordingly, flooding water at high temperature heated by the after-heat of the reactor core is returned again into the flooding gap to cool the reactor core repeatedly. (T.M.)

  17. Development of metallic high integrity containers

    International Nuclear Information System (INIS)

    Temus, C.J.; Porter, S.A.; Kent, J.D.; Goetsch, S.D.

    1985-01-01

    This paper presents the development program for metallic high integrity containers (HIC's). The need for a high strength, thin walled HIC became apparent with the implementation of 10 CFR 61 in late 1983. The existing containers that were in use at that time were made of either low strength material (polyethylene) or bulky, heavy material (concrete). Neither of these materials met the need for a high strength, volume and weight efficient container that could survive the deep burial environment of sites such as Hanford, Washington. Various alloys were considered for corrosion resistance for a 300 year life, high strength and toughness, and elastic stability to meet the requirements of 10 CFR 61. The alloy allowed for great flexibility in design to accommodate various waste forms. The containers were developed in various sizes with several different closures designed to minimize operator exposure during the loading operation. These design features provide the industry with efficient, disposable packages for a wide variety of waste forms. The paper describes the analytical methodology and prototype test program. The analytical methods included finite element modeling of the burial conditions, prediction of drop performance,and elastic stability analysis. Prototype testing included leak tests and drop tests at various container orientations from heights of up to 25 feet

  18. NPP Krsko Severe Accident Management Guidelines Implementation

    International Nuclear Information System (INIS)

    Basic, I.; Krajnc, B.; Bilic-Zabric, T.; Spiler, J.

    2002-01-01

    Severe Accident Management is a framework to identify and implement the Emergency Response Capabilities that can be used to prevent or mitigate severe accidents and their consequences. The USA NRC has indicated that the development of a licensee plant specific accident management program will be required in order to close out the severe accident regulatory issue (Ref. SECY-88-147). Generic Letter 88-20 ties the Accident management Program to IPE for each plant. The SECY-89-012 defines those actions taken during the course of an accident by the plant operating and technical staff to: 1) prevent core damage, 2) terminate the progress of core damage if it begins and retain the core within the reactor vessel, 3) maintain containment integrity as long as possible, and 4) minimize offsite releases. The subject of this paper is to document the severe accident management activities, which resulted in a plant specific Severe Accident Management Guidelines implementation. They have been developed based on the Krsko IPE (Individual Plant Examination) insights, Generic WOG SAMGs (Westinghouse Owners Group Severe Accident Management Guidances) and plant specific documents developed within this effort. Among the required plant specific actions the following are the most important ones: Identification and documentation of those Krsko plant specific severe accident management features (which also resulted from the IPE investigations). The development of the Krsko plant specific background documents (Severe Accident Plant Specific Strategies and SAMG Setpoint Calculation). Also, paper discusses effort done in the areas of NPP Krsko SAMG review (internal and external ), validation on Krsko Full Scope Simulator (Severe Accident sequences are simulated by MAAP4 in real time) and world 1st IAEA Review of Accident Management Programmes (RAMP). (author)

  19. Improved hydrogen distribution calculation in the containment using the coupled codes MELCOR and GASFLOW for the analysis of severe accidents in nuclear power plants; Verbesserte Berechnung der Wasserstoffverteilung im Sicherheitsbehaelter bei der Analyse schwerer Stoerfaelle in Kernkraftwerken durch Kopplung von MELCOR und GASFLOW

    Energy Technology Data Exchange (ETDEWEB)

    Szabo, Tobias

    2014-09-01

    The risk of a hydrogen combustion within a containment of a pressurized water reactor during a severe loss of coolant accident (LOCA) is evaluated using numerical simulations. The code MELCOR provides integral analysis capabilities for severe accidents. Yet, its Lumped Parameter (LP) model provides less accurate information about on thermal hydraulics within the containment during a LOCA. GASFLOW is a CFD code that simulates both the local hydrogen distribution and the pressure inside the containment more realistically. Currently, to perform these GASFLOW simulations, the common procedure is to use a source term from a previous MELCOR calculation. However, with this approach, the influence of the more realistic GASFLOW pressure on the mass flow through the leak cannot be taken into account. This inconsistency is overcome by coupling both codes in this thesis. Here, the MELCOR instance is responsible for the primary and secondary systems. At the same time, the GASFLOW instance predicts the thermal hydraulics of the containment. The more accurate containment pressure from the GASFLOW instance is used in the MELCOR instance to calculate consistent outflow rates through the leak. In order to couple both codes, the existing interface in MELCOR is modified and a new interface for GASFLOW is developed and implemented. To begin with, the hydrogen distribution inside a generic containment is calculated by MELCOR using a typical coarse LP nodalization and a refined one. The results obtained are compared to a GASFLOW simulation. It is shown that the refinement only leads to a better agreement with the GASFLOW result if the correct flow directions are predefined by the nodalization. The safety relevant, local peak concentrations of hydrogen cannot be resolved by MELCOR. Consequently, the use of the CFD code is indispensable. The correct functioning of the coupling is proven within four steps. At first, the modified MELCOR interface is checked by computing a test case using two

  20. 7 CFR 201.42 - Small containers.

    Science.gov (United States)

    2010-01-01

    ... 7 Agriculture 3 2010-01-01 2010-01-01 false Small containers. 201.42 Section 201.42 Agriculture... REGULATIONS Sampling in the Administration of the Act § 201.42 Small containers. In sampling seed in small containers that it is not practical to sample as required in § 201.41, a portion of one unopened container or...

  1. Optimizing yard operations in port container terminals

    DEFF Research Database (Denmark)

    Kallehauge, Louise Sibbesen

    2005-01-01

    This paper deals with the problem of positioning containers in a yard block of a port container terminal. The objective of the container positioning problem (CPP) is to minimise the total handling time in the block, i.e. the time required for storage and reshuffling of containers. One...

  2. 46 CFR 160.036-6 - Container.

    Science.gov (United States)

    2010-10-01

    ... 46 Shipping 6 2010-10-01 2010-10-01 false Container. 160.036-6 Section 160.036-6 Shipping COAST... § 160.036-6 Container. (a) General. The container for storing the signals on lifeboats and liferafts is not required to be of a special design or be approved by the Coast Guard. The container must meet the...

  3. 7 CFR 906.340 - Container, pack, and container marking regulations.

    Science.gov (United States)

    2010-01-01

    ... 7 Agriculture 8 2010-01-01 2010-01-01 false Container, pack, and container marking regulations... AGRICULTURE ORANGES AND GRAPEFRUIT GROWN IN LOWER RIO GRANDE VALLEY IN TEXAS Container and Pack Requirements § 906.340 Container, pack, and container marking regulations. (a) No handler shall handle any variety of...

  4. Waste container and method for containing waste

    International Nuclear Information System (INIS)

    Ono, Akira; Matsushita, Mitsuhiro; Doi, Makoto; Nakatani, Seiichi.

    1990-01-01

    In a waste container, water-proof membranes and rare earth element layers are formed on the inner surface of a steel plate concrete container in which steel plates are embedded. Further, rear earth element detectors are disposed each from the inner side of the steel plate concrete container by way of a pressure pipe to the outer side of the container. As a method for actually containing wastes, when a plurality of vessels in which wastes are fixed are collectively enhoused to the waste container, cussioning materials are attached to the inner surface of the container and wastes fixing containers are stacked successively in a plurality of rows in a bag made of elastic materials. Subsequently, fixing materials are filled and tightly sealed in the waste container. When the waste container thus constituted is buried underground, even if it should be deformed to cause intrusion of rain water to the inside of the container, the rare earth elements in the container dissolved in the rain water can be detected by the detectors, the containers are exchanged before the rain water intruding to the inner side is leached to the surrounding ground, to previously prevent the leakage of radioactive nuclides. (K.M.)

  5. [Favorable current prognosis after HLA-identical bone marrow transplantation for children with required severe aplastic anemia; evaluation of 30 years of bone marrow transplantation at the Leiden University Medical Center

    NARCIS (Netherlands)

    Steekelenburg, M. van; Weel-Sipman, M.H. van; Zwinderman, A.H.; Hoogerbrugge, P.M.; Vossen, J.M.J.J.; Egeler, R.M.

    2002-01-01

    OBJECTIVE: To evaluate the results of 30 years of allogeneic HLA-identical bone marrow transplantation (BMT) as the treatment for children with acquired severe aplastic anaemia. DESIGN: Retrospective, descriptive. METHOD: Of all patients who underwent an HLA-identical sibling-donor BMT for severe

  6. Allergy in severe asthma.

    Science.gov (United States)

    Del Giacco, S R; Bakirtas, A; Bel, E; Custovic, A; Diamant, Z; Hamelmann, E; Heffler, E; Kalayci, Ö; Saglani, S; Sergejeva, S; Seys, S; Simpson, A; Bjermer, L

    2017-02-01

    It is well recognized that atopic sensitization is an important risk factor for asthma, both in adults and in children. However, the role of allergy in severe asthma is still under debate. The term 'Severe Asthma' encompasses a highly heterogeneous group of patients who require treatment on steps 4-5 of GINA guidelines to prevent their asthma from becoming 'uncontrolled', or whose disease remains 'uncontrolled' despite this therapy. Epidemiological studies on emergency room visits and hospital admissions for asthma suggest the important role of allergy in asthma exacerbations. In addition, allergic asthma in childhood is often associated with severe asthma in adulthood. A strong association exists between asthma exacerbations and respiratory viral infections, and interaction between viruses and allergy further increases the risk of asthma exacerbations. Furthermore, fungal allergy has been shown to play an important role in severe asthma. Other contributing factors include smoking, pollution and work-related exposures. The 'Allergy and Asthma Severity' EAACI Task Force examined the current evidence and produced this position document on the role of allergy in severe asthma. © 2016 John Wiley & Sons A/S. Published by John Wiley & Sons Ltd.

  7. Accident management for severe accidents

    International Nuclear Information System (INIS)

    Bari, R.A.; Pratt, W.T.; Lehner, J.; Leonard, M.; Disalvo, R.; Sheron, B.

    1988-01-01

    The management of severe accidents in light water reactors is receiving much attention in several countries. The reduction of risk by measures and/or actions that would affect the behavior of a severe accident is discussed. The research program that is being conducted by the US Nuclear Regulatory Commission focuses on both in-vessel accident management and containment and release accident management. The key issues and approaches taken in this program are summarized. 6 refs

  8. Container crane for sea freight containers

    NARCIS (Netherlands)

    Luttekes, E.; Rijsenbrij, J.C.

    2001-01-01

    The invention relates to a container crane for loading and unloading seaborne containers. The container crane comprises a bridge girder (7), a jib (8), at least two crabs (11, 12) which can travel along the said bridge girder and/or jib and are provided with hoist means for lifting and lowering the

  9. Continuous containment monitoring with containment pressure fluctuation

    International Nuclear Information System (INIS)

    Dick, J.E.

    1996-01-01

    The monitoring of the integrity of containments particularly but not exclusively for nuclear plants is dealt with in this invention. While this application is primarily concerned with containment monitoring in the context of the single unit design, it is expected that the concepts presented will be universally applicable to any containment design, including containments for non-nuclear applications such as biological laboratories. The nuclear industry has long been interested in a means of monitoring containment integrity on a continuous basis, that is, while the reactor is operating normally. 12 refs., 2 figs

  10. Containment Sodium Chemistry Models in MELCOR.

    Energy Technology Data Exchange (ETDEWEB)

    Louie, David [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Humphries, Larry L. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Denman, Matthew R [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States)

    2017-04-01

    To meet regulatory needs for sodium fast reactors’ future development, including licensing requirements, Sandia National Laboratories is modernizing MELCOR, a severe accident analysis computer code developed for the U.S. Nuclear Regulatory Commission (NRC). Specifically, Sandia is modernizing MELCOR to include the capability to model sodium reactors. However, Sandia’s modernization effort primarily focuses on the containment response aspects of the sodium reactor accidents. Sandia began modernizing MELCOR in 2013 to allow a sodium coolant, rather than water, for conventional light water reactors. In the past three years, Sandia has been implementing the sodium chemistry containment models in CONTAIN-LMR, a legacy NRC code, into MELCOR. These chemistry models include spray fire, pool fire and atmosphere chemistry models. Only the first two chemistry models have been implemented though it is intended to implement all these models into MELCOR. A new package called “NAC” has been created to manage the sodium chemistry model more efficiently. In 2017 Sandia began validating the implemented models in MELCOR by simulating available experiments. The CONTAIN-LMR sodium models include sodium atmosphere chemistry and sodium-concrete interaction models. This paper presents sodium property models, the implemented models, implementation issues, and a path towards validation against existing experimental data.

  11. Severe childhood malnutrition.

    Science.gov (United States)

    Bhutta, Zulfiqar A; Berkley, James A; Bandsma, Robert H J; Kerac, Marko; Trehan, Indi; Briend, André

    2017-09-21

    The main forms of childhood malnutrition occur predominantly in children malnutrition. Here, we use the term 'severe malnutrition' to describe these conditions to better reflect the contributions of chronic poverty, poor living conditions with pervasive deficits in sanitation and hygiene, a high prevalence of infectious diseases and environmental insults, food insecurity, poor maternal and fetal nutritional status and suboptimal nutritional intake in infancy and early childhood. Children with severe malnutrition have an increased risk of serious illness and death, primarily from acute infectious diseases. International growth standards are used for the diagnosis of severe malnutrition and provide therapeutic end points. The early detection of severe wasting and kwashiorkor and outpatient therapy for these conditions using ready-to-use therapeutic foods form the cornerstone of modern therapy, and only a small percentage of children require inpatient care. However, the normalization of physiological and metabolic functions in children with malnutrition is challenging, and children remain at high risk of relapse and death. Further research is urgently needed to improve our understanding of the pathophysiology of severe malnutrition, especially the mechanisms causing kwashiorkor, and to develop new interventions for prevention and treatment.

  12. Containment and release management

    International Nuclear Information System (INIS)

    Lehner, J.R.; Pratt, W.T.

    1988-01-01

    Reducing the risk from potentially severe accidents by appropriate accident management strategies is receiving increased attention from the international reactor safety community. Considerable uncertainty still surrounds some of the physical phenomena likely to occur during a severe accident. The USNRC, in developing its research plan for accident management, wants to ensure that both the developers and implementers of accident management strategies are aware of the uncertainty associated with the plant operators' ability to correctly diagnose an accident, as well as the uncertainties associated with various preventive and mitigative strategies. The use of a particular accident management strategy can have both positive and negative effects on the status of a plant and these effects must be carefully weighed before a particular course of action is chosen and implemented. By using examples of severe accident scenarios, initial insights are presented here regarding the indications plant operators may have to alert them to particular accident states. Insights are also offered on the various management actions operators and plant technical staff might pursue for particular accident situations and the pros and cons associated with such actions. The examples given are taken for the most part from the containment and release phase of accident management, since this is the current focus of the effort in the accident management area at Brookhaven National Laboratory. 2 refs

  13. Containment integrity analysis under accidents

    International Nuclear Information System (INIS)

    Lin Chengge; Zhao Ruichang; Liu Zhitao

    2010-01-01

    Containment integrity analyses for current nuclear power plants (NPPs) mainly focus on the internal pressure caused by design basis accidents (DBAs). In addition to the analyses of containment pressure response caused by DBAs, the behavior of containment during severe accidents (SAs) are also evaluated for AP1000 NPP. Since the conservatism remains in the assumptions,boundary conditions and codes, margin of the results of containment integrity analyses may be overestimated. Along with the improvements of the knowledge to the phenomena and process of relevant accidents, the margin overrated can be appropriately reduced by using the best estimate codes combined with the uncertainty methods, which could be beneficial to the containment design and construction of large passive plants (LPP) in China. (authors)

  14. Severe service sealing solutions

    International Nuclear Information System (INIS)

    Metcalfe, R.; Wensel, R.

    1994-09-01

    Successful sealing usually requires much more than initial leak-tightness. Friction and wear must also be acceptable, requiring a good understanding of tribology at the sealing interface. This paper describes various sealing solutions for severe service conditions. The CAN2A and CAN8 rotary face seals use tungsten carbide against carbon-graphite to achieve low leakage and long lifetime in nuclear main coolant pumps. The smaller CAN6 seal successfully uses tungsten carbide against silicon carbide in reactor water cleanup pump service. Where friction in CANDU fuelling machine rams must be essentially zero, a hydrostatic seal using two silicon carbide faces is the solution. In the NRU reactor moderator pumps, where pressure is much lower, eccentric seals that prevent boiling at the seal faces are giving excellent service. All these rotary face seals rely on supplementary elastomer seals between their parts. An integrated engineering approach to high performance sealing with O-rings is described. This is epitomized in critical Space Shuttle applications, but is increasingly being applied in CANDU plants. It includes gland design, selection and qualification of material, quality assurance, detection of defects and the effects of lubrication, surface finish, squeeze, stretch and volume constraints. In conclusion, for the severe service applications described, customized solutions have more than paid for themselves by higher reliability, lower maintenance requirements and reduced outage time. (author)

  15. APR1400 Containment Simulation with CONTAIN code

    Energy Technology Data Exchange (ETDEWEB)

    Hwang, Moon Kyu; Chung, Bub Dong [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2010-05-15

    The more realistic containment pressure variation predicted by the CONTAIN code through the coupled analysis during a large break loss of coolant accident in the nuclear power plant is expected to provide more accurate prediction for the plant behavior than a standalone MARS-KS calculation. The input deck has been generated based on the already available ARP- 1400 input for CONTEMPT code. Similarly to the CONTEMPT input deck, a simple two-cell model was adopted to model the containment behavior, one cell for the containment inner volume and another cell for the environment condition. The developed input for the CONTAIN code is to be eventually applied for the coupled code calculation of MARS-KS/CONTAIN

  16. APR1400 Containment Simulation with CONTAIN code

    International Nuclear Information System (INIS)

    Hwang, Moon Kyu; Chung, Bub Dong

    2010-01-01

    The more realistic containment pressure variation predicted by the CONTAIN code through the coupled analysis during a large break loss of coolant accident in the nuclear power plant is expected to provide more accurate prediction for the plant behavior than a standalone MARS-KS calculation. The input deck has been generated based on the already available ARP- 1400 input for CONTEMPT code. Similarly to the CONTEMPT input deck, a simple two-cell model was adopted to model the containment behavior, one cell for the containment inner volume and another cell for the environment condition. The developed input for the CONTAIN code is to be eventually applied for the coupled code calculation of MARS-KS/CONTAIN

  17. Asbestos-Containing Materials (ACM) and Demolition

    Science.gov (United States)

    There are specific federal regulatory requirements that require the identification of asbestos-containing materials (ACM) in many of the residential buildings that are being demolished or renovated by a municipality.

  18. ANS severe accident program overview & planning document

    Energy Technology Data Exchange (ETDEWEB)

    Taleyarkhan, R.P.

    1995-09-01

    The Advanced Neutron Source (ANS) severe accident document was developed to provide a concise and coherent mechanism for presenting the ANS SAP goals, a strategy satisfying these goals, a succinct summary of the work done to date, and what needs to be done in the future to ensure timely licensability. Guidance was received from various bodies [viz., panel members of the ANS severe accident workshop and safety review committee, Department of Energy (DOE) orders, Nuclear Regulatory Commission (NRC) requirements for ALWRs and advanced reactors, ACRS comments, world-wide trends] were utilized to set up the ANS-relevant SAS goals and strategy. An in-containment worker protection goal was also set up to account for the routine experimenters and other workers within containment. The strategy for achieving the goals is centered upon closing the severe accident issues that have the potential for becoming certification issues when assessed against realistic bounding events. Realistic bounding events are defined as events with an occurrency frequency greater than 10{sup {minus}6}/y. Currently, based upon the level-1 probabilistic risk assessment studies, the realistic bounding events for application for issue closure are flow blockage of fuel element coolant channels, and rapid depressurization-related accidents.

  19. Management of severe accidents

    International Nuclear Information System (INIS)

    Jankowski, M.W.

    1987-01-01

    The definition and the multidimensionality aspects of accident management have been reviewed. The suggested elements in the development of a programme for severe accident management have been identified and discussed. The strategies concentrate on the two tiered approaches. Operative management utilizes the plant's equipment and operators capabilities. The recovery managment concevtrates on preserving the containment, or delaying its failure, inhibiting the release, and on strategies once there has been a release. The inspiration for this paper was an excellent overview report on perspectives on managing severe accidents in commercial nuclear power plants and extending plant operating procedures into the severe accident regime; and by the most recent publication of the International Nuclear Safety Advisory Group (INSAG) considering the question of risk reduction and source term reduction through accident prevention, management and mitigation. The latter document concludes that 'active development of accident management measures by plant personnel can lead to very large reductions in source terms and risk', and goes further in considering and formulating the key issue: 'The most fruitful path to follow in reducing risk even further is through the planning of accident management.' (author)

  20. Management of severe accidents

    International Nuclear Information System (INIS)

    Jankowski, M.W.

    1988-01-01

    The definition and the multidimensionality aspects of accident management have been reviewed. The suggested elements in the development of a programme for severe accident management have been identified and discussed. The strategies concentrate on the two tiered approaches. Operative management utilizes the plant's equipment and operators capabilities. The recovery management concentrates on preserving the containment, or delaying its failure, inhibiting the release, and on strategies once there has been a release. The inspiration for this paper was an excellent overview report on perspectives on managing severe accidents in commercial nuclear power plants and extending plant operating procedures into the severe accident regime; and by the most recent publication of the International Nuclear Safety Advisory Group (INSAG) considering the question of risk reduction and source term reduction through accident prevention, management and mitigation. The latter document concludes that active development of accident management measures by plant personnel can lead to very large reductions in source terms and risk, and goes further in considering and formulating the key issue: The most fruitful path to follow in reducing risk even further is through the planning of accident management

  1. Conceptual study on the containment design aiming at 'no evacuation'

    International Nuclear Information System (INIS)

    Andou, Kouji; Takii, Taichi; Kikuyama, Tomohiko; Taminami, Tatsuya

    2003-01-01

    The next generation reactors represented in ABWR-II should enhance not only economics but also safety. Especially, the ideal target of 'No Evacuation', that is, no FP (Fission Product) release in severe accidents should be required. This paper provides the conceptual design achieving 'No Evacuation' by using only the passive systems, that are, the passive containment cooling system (PCCS), the large amount of water inside containment, outside pool by utilizing gap between containment and surrounding building, and the natural heat removal from the containment surface to atmosphere. Furthermore, it is also easy to adopt the countermeasure for airplane crash by using a dome shelter and the dispersed layout as an option. At the same time, the amount of the construction material of this concept is competitive comparing with that of the conventional BWR because it is easy to use the steel structure or the steel plate reinforced concrete structure over a wide area. (author)

  2. Elements of thought on corium containment strategy in reactor vessel

    International Nuclear Information System (INIS)

    2015-01-01

    As accidents with core fusion are taken into account for the design of third-generation nuclear reactors, this brief document presents the corium containment strategy for a reactor vessel, its limitations, as well as research programs undertaken by the IRSN in this field. The report describes the controlled management of a severe accident, the major objective being to minimise releases in the environment, that which requires to maintain the reactor containment enclosure tightness. Practical actions are briefly indicated. Key points indicating the feasibility of a strategy of containment in vessel are discussed. The impact of reactor power on the robustness of an approach with containment in vessel is also discussed. An overview of technological evolutions and contributions of researches made by the IRSN is finally proposed

  3. Empty Container Logistics

    Directory of Open Access Journals (Sweden)

    Jakov Karmelić

    2012-05-01

    Full Text Available Within the whole world container traffic, the largest share of containers is in the status of repositioning. Container repositioning results from the need for harmonization between the point of empty container accumulation and the point of demand, and waiting time for the availability of the first next transport of cargo. This status of containers on the container market is the consequence of imbalances in the worldwide trade distribution on most important shipping routes. The need for fast and effective reallocation of empty containers causes high costs and often represents an obstacle affecting the efficiency of port container terminals and inland carriers.In accordance with the above issue, this paper is mainly focused on the analysis of the data concerning global container capacities and the roots of container equipment imbalances, with the aim of determining the importance of empty container management and the need for empty container micro-logistic planning at the spread port area.

  4. Definition of containment issues

    International Nuclear Information System (INIS)

    Walker, D.H.

    1982-01-01

    Public Law 96-567 Nuclear Safety Research, Development and Demonstration Act of 1980, directed the US Department of Energy (DOE) to provide an accelerated and coordinated program for developing practical generic improvements that would enhance the capability for safe, reliable and economical operation of Light Water Nuclear Reactor Power Stations. The DOE approach to defining such a program will consist of two phases, (1) definition of program requirements and (2) implementation of the program plan. This paper summarizes the results of the program definition phase for the containment integrity function. The definition phase effort was carried out by two groups of knowledgeable technical experts from the nuclear industry, one of which addressed containment integrity. Tabulated in the paper are the issues identified by the working groups and their associated priorities. Also tabulated are those high priority issues for which ongoing programs do not appear to provide sufficient information to resolve the issue. The results of this review show that existing programs to a great extent address existing issues in a manner such that the issues should be resolved by the programs

  5. Passive heat transport in advanced CANDU containment

    International Nuclear Information System (INIS)

    Krause, M.; Mathew, P.M.

    1994-01-01

    A passive CANDU containment design has been proposed to provide the necessary heat removal following a postulated accident to maintain containment integrity. To study its feasibility and to optimize the design, multi-dimensional containment modelling may be required. This paper presents a comparison of two CFD codes, GOTHIC and PHOENICS, for multi-dimensional containment analysis and gives pressure transient predictions from a lumped-parameter and a three-dimensional GOTHIC model for a modified CANDU-3 containment. GOTHIC proved suitable for multidimensional post-accident containment analysis, as shown by the good agreement with pressure transient predictions from PHOENICS. GOTHIC is, therefore, recommended for passive CANDU containment modelling. (author)

  6. Empty Container Logistics

    OpenAIRE

    Jakov Karmelić; Čedomir Dundović; Ines Kolanović

    2012-01-01

    Within the whole world container traffic, the largest share of containers is in the status of repositioning. Container repositioning results from the need for harmonization between the point of empty container accumulation and the point of demand, and waiting time for the availability of the first next transport of cargo. This status of containers on the container market is the consequence of imbalances in the worldwide trade distribution on most important shipping routes. The need for fast a...

  7. Alternative containment integrity test methods, an overview of possible techniques

    International Nuclear Information System (INIS)

    Spletzer, B.L.

    1986-01-01

    A study is being conducted to develop and analyze alternative methods for testing of containment integrity. The study is focused on techniques for continuously monitoring containment integrity to provide rapid detection of existing leaks, thus providing greater certainty of the integrity of the containment at any time. The study is also intended to develop techniques applicable to the currently required Type A integrated leakage rate tests. A brief discussion of the range of alternative methods currently being considered is presented. The methods include applicability to all major containment types, operating and shutdown plant conditions, and quantitative and qualitative leakage measurements. The techniques are analyzed in accordance with the current state of knowledge of each method. The bulk of the techniques discussed are in the conceptual stage, have not been tested in actual plant conditions, and are presented here as a possible future direction for evaluating containment integrity. Of the methods considered, no single method provides optimum performance for all containment types. Several methods are limited in the types of containment for which they are applicable. The results of the study to date indicate that techniques for continuous monitoring of containment integrity exist for many plants and may be implemented at modest cost

  8. Accomplishments and challenges of the severe accident research

    International Nuclear Information System (INIS)

    Sehga, B.R.

    1998-01-01

    This paper describes the progress of the severe accident research since 1980, in terms of the accomplishments made so far and the challenges that remain. Much has been accomplished: many important safety issues have been resolved and consensus is near on some others. However, some of the previously identified safety issues remain as challenges, while some new ones have arisen due to the shift in focus from containment integrity to vessel integrity. New reactor designs have also created some new challenges. In general, the regulatory demands in new reactor designs are much stricter, thereby requiring much greater attention to the safety issues concerned with the containment design of the new large reactors

  9. Nuclear reactor container

    International Nuclear Information System (INIS)

    Ishiyama, Takenori.

    1989-01-01

    This invention concerns a nuclear reactor container in which heat is removed from a container by external water injection. Heat is removed from the container by immersing the lower portion of the container into water and scattering spary water from above. Thus, the container can be cooled by the spray water falling down along the outer wall of the container to condensate and cool vapors filled in the container upon occurrence of accidents. Further, since the inside of the container can be cooled also during usual operation, it can also serve as a dry well cooler. Accordingly, heat is removed from the reactor container upon occurrence of accidents by the automatic operation of a spray device corresponding to the change of the internal temperature and the pressure in the reactor container. Further, since all of these devices are disposed out of container, maintenance is also facilitated. (I.S.)

  10. From containment to ... syzygy

    Energy Technology Data Exchange (ETDEWEB)

    Carr, H.M.

    1995-04-18

    The U S. is at an historic crossroads following the end of the Cold War. The old twin themes of containment and deterrence must now give way to a newer vision of the U.S. role as we approach the 21st century. This paper follows a visioning process requiring development of alternatives based on signposts, values and frameworks. Signposts are current domestic and global environments revealing a U.S. in economic trouble with budget and trade defidts, a falling dollar and multiplying peace operations at a time when Europe and Japan are becoming economic superpowers. Although U.S. values must be protected, economic competition requires increased emphasis on realpolitik. A balanced framework of internationalism and reduced multilateralism will suit the current environment and U.S. purposes. Recognizing that the U.S. must retain leadership to protect national interests, the vision unfolds as an alignment of major powers-a syzygy of purpose-with Europe, Japan, and the U.S. in a concert of power, sharing economic, political and military burdens to ensure world stability. Thus, the new U.S. role could be primus inter pares of a Pax Consortis with common interests and goals, allowing the U.S. time to restore its economic vitality.

  11. Accomplishments and challenges of the severe accident research

    International Nuclear Information System (INIS)

    Sehgal, B.R.

    2001-01-01

    This paper briefly describes the progress of the severe accident research since 1980, in terms of the accomplishments made so far and the challenges that remain. Much has been accomplished: many important safety issues have been resolved and consensus is near on some others. However, some of the previously identified safety issues remain as challenges, while some new ones have arisen due to the shift in focus from containment to vessel integrity. New reactor designs have also created some new challenges. In general, the regulatory demands for new reactor designs are stricter, thereby requiring much greater attention to the safety issues concerned with the containment design of the new large reactors, and to the accident management procedures for mitigating the consequences of a severe accident. We apologize for not providing references to many fine investigations that contributed to the great progress made so far in the severe accident research

  12. Proceedings of the third international conference on containment design and operation. v.1

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1994-12-31

    The second international conference on containment design and operation included sessions on the following topics: performance and regulatory requirements; radionuclide behaviour; severe accident design and analysis; operation, maintenance, leaking and aging of containment systems; thermal hydraulic behaviour of containment systems; hydrogen mixing and mitigation; design methods and concepts; code validation; structural analysis and response tests; passive safety systems; aerosol behaviour; containment reliability, integrity, and risk assessment; hydrogen deflagration and detonation. Due prominence was given to CANDU and other PHWR reactors. The individual papers have been abstracted separately.

  13. Proceedings of the third international conference on containment design and operation. v.1

    International Nuclear Information System (INIS)

    1994-01-01

    The second international conference on containment design and operation included sessions on the following topics: performance and regulatory requirements; radionuclide behaviour; severe accident design and analysis; operation, maintenance, leaking and aging of containment systems; thermal hydraulic behaviour of containment systems; hydrogen mixing and mitigation; design methods and concepts; code validation; structural analysis and response tests; passive safety systems; aerosol behaviour; containment reliability, integrity, and risk assessment; hydrogen deflagration and detonation. Due prominence was given to CANDU and other PHWR reactors. The individual papers have been abstracted separately

  14. 27 CFR 44.254 - Shipping containers.

    Science.gov (United States)

    2010-04-01

    ... 27 Alcohol, Tobacco Products and Firearms 2 2010-04-01 2010-04-01 false Shipping containers. 44.254 Section 44.254 Alcohol, Tobacco Products and Firearms ALCOHOL AND TOBACCO TAX AND TRADE BUREAU... Requirements § 44.254 Shipping containers. Each shipping case, crate, or other container, in which cigars are...

  15. Materials for high-level waste containment

    International Nuclear Information System (INIS)

    Marsh, G.P.

    1982-01-01

    The function of the high-level radioactive waste container in storage and of a container/overpack combination in disposal is considered. The consequent properties required from potential fabrication materials are discussed. The strategy adopted in selecting containment materials and the experimental programme underway to evaluate them are described. (U.K.)

  16. 46 CFR 160.037-6 - Container.

    Science.gov (United States)

    2010-10-01

    ... 46 Shipping 6 2010-10-01 2010-10-01 false Container. 160.037-6 Section 160.037-6 Shipping COAST...: SPECIFICATIONS AND APPROVAL LIFESAVING EQUIPMENT Hand Orange Smoke Distress Signals § 160.037-6 Container. (a) General. The container for storing the signals on lifeboats and liferafts is not required to be of a...

  17. Alternative pathways to antimatter containment

    International Nuclear Information System (INIS)

    Rejcek, J.M.; Browder, M.K.; Fry, J.L.; Koymen, A.; Weiss, A.H.

    2003-01-01

    Antimatter containment is a gateway technology for future advancements in many areas. Immediate applications in propulsion, medicine, and instrumentation have already been envisioned and many others are yet to be considered. Key to this technological advance is identifying one or more pathways to achieve safe reliable containment of antimatter in sufficient quantities to be useful on an engineering and industrial scale. The goal of this paper is to review current approaches and discuss possible alternative pathways to antimatter containment. Specifically, this paper will address the possibility of designing a solid-state containment system that will safely hold antimatter in quantities dense enough to be of any engineering utility. A discussion of the current research, the needed engineering requirements, and a survey of current research is presented

  18. Dyslexia and Severe Reading Disability.

    Science.gov (United States)

    Ngandu, Kathleen M.

    This handbook contains advice for the teacher in diagnosing dyslexia and developing an individualized program for overcoming severe reading problems. Observable characteristics of dyslexia are listed as an aid to the teacher's diagnosis, but it is emphasized that cooperation between the teacher and a reading specialist is of great importance in…

  19. Epidemiology of severe trauma.

    Science.gov (United States)

    Alberdi, F; García, I; Atutxa, L; Zabarte, M

    2014-12-01

    Major injury is the sixth leading cause of death worldwide. Among those under 35 years of age, it is the leading cause of death and disability. Traffic accidents alone are the main cause, fundamentally in low- and middle-income countries. Patients over 65 years of age are an increasingly affected group. For similar levels of injury, these patients have twice the mortality rate of young individuals, due to the existence of important comorbidities and associated treatments, and are more likely to die of medical complications late during hospital admission. No worldwide, standardized definitions exist for documenting, reporting and comparing data on severely injured trauma patients. The most common trauma scores are the Abbreviated Injury Scale (AIS), the Injury Severity Score (ISS) and the Trauma and Injury severity Score (TRISS). Documenting the burden of injury also requires evaluation of the impact of post-trauma impairments, disabilities and handicaps. Trauma epidemiology helps define health service and research priorities, contributes to identify disadvantaged groups, and also facilitates the elaboration of comparable measures for outcome predictions. Copyright © 2014 Elsevier España, S.L.U. y SEMICYUC. All rights reserved.

  20. Rapid Sampling from Sealed Containers

    International Nuclear Information System (INIS)

    Johnston, R.G.; Garcia, A.R.E.; Martinez, R.K.; Baca, E.T.

    1999-01-01

    The authors have developed several different types of tools for sampling from sealed containers. These tools allow the user to rapidly drill into a closed container, extract a sample of its contents (gas, liquid, or free-flowing powder), and permanently reseal the point of entry. This is accomplished without exposing the user or the environment to the container contents, even while drilling. The entire process is completed in less than 15 seconds for a 55 gallon drum. Almost any kind of container can be sampled (regardless of the materials) with wall thicknesses up to 1.3 cm and internal pressures up to 8 atm. Samples can be taken from the top, sides, or bottom of a container. The sampling tools are inexpensive, small, and easy to use. They work with any battery-powered hand drill. This allows considerable safety, speed, flexibility, and maneuverability. The tools also permit the user to rapidly attach plumbing, a pressure relief valve, alarms, or other instrumentation to a container. Possible applications include drum venting, liquid transfer, container flushing, waste characterization, monitoring, sampling for archival or quality control purposes, emergency sampling by rapid response teams, counter-terrorism, non-proliferation and treaty verification, and use by law enforcement personnel during drug or environmental raids

  1. Underground storage tanks containing hazardous chemicals

    International Nuclear Information System (INIS)

    Wise, R.F.; Starr, J.W.; Maresca, J.W. Jr.; Hillger, R.W.; Tafuri, A.N.

    1991-01-01

    The regulations issued by the United States Environmental Protection Agency in 1988 require, with several exceptions, that underground storage tank systems containing petroleum fuels and hazardous chemicals be routinely tested for releases. This paper summarizes the release detection regulations for tank systems containing chemicals and gives a preliminary assessment of the approaches to release detection currently being used. To make this assessment, detailed discussions were conducted with providers and manufacturers of leak detection equipment and testing services, owners or operators of different types of chemical storage tank systems, and state and local regulators. While these discussions were limited to a small percentage of each type of organization, certain observations are sufficiently distinctive and important that they are reported for further investigation and evaluation. To make it clearer why certain approaches are being used, this paper also summarizes the types of chemicals being stored, the effectiveness of several leak detection testing systems, and the number and characteristics of the tank systems being used to store these products

  2. Passive heat removal from containment

    International Nuclear Information System (INIS)

    Gou, P.F.; Townsend, H.E.

    1990-01-01

    This patent describes a heat removal system for removing heat from a containment of a nuclear reactor. It comprises: a sealed suppression chamber in the containment; means for venting steam from the nuclear reactor into the suppression chamber upon occurrence of an event requiring dissipation of heat from the nuclear reactor. The suppression chamber containing a quantity of water; the suppression chamber having a gas-containing space above the water; a heat exchanger disposed within the gas-containing space of the suppression chamber; the heat exchanger including an enclosed structure for holding a heat-exchange fluid; means for metering a supply of heat-exchange fluid to the heat exchanger to maintain a predetermined level thereof in the enclosed structure. The heat-exchange fluid boiling in the heat exchanger in consequence of heat transfer thereto from steam present in the suppression chamber; means for separating a heat-exchange fluid vapor in the heat exchanger from the heat-exchange fluid; and means for discharging the vapor immediately following its separation from heat-exchange fluid directly from the heat exchanger to a location exterior of the containment, whereby heat is discharged from the suppression chamber, and the containment is maintained at a temperature and pressure below its design value

  3. Emergency reactor container cooling facility

    International Nuclear Information System (INIS)

    Suzuki, Hiroaki; Matsumoto, Tomoyuki.

    1992-01-01

    The present invention concerns an emergency cooling facility for a nuclear reactor container having a pressure suppression chamber, in which water in the suppression chamber is effectively used for cooling the reactor container. That is, the lower portion of a water pool in the pressure suppression chamber and the inside of the reactor container are connected by a pipeline. The lower end of the pipeline and a pressurized incombustible gas tank disposed to the outside of the reactor container are connected by a pipeline by way of valves. Then, when the temperature of the lower end of the pressure vessel exceeds a predetermined value, the valves are opened. If the valves are opened, the incombustible gas flows into the lower end of the pipeline connecting the lower portion of the water pool in the pressure suppression chamber and the inside of the reactor container. Since the inside of the pipeline is a two phase flow comprising a mixture of a gas phase and a liquid phase, the average density is decreased. Therefore, the water level of the two phase flow is risen by the level difference between the inside and the outside of the pipeline and, finally, the two phase mixture is released into the reactor container. As a result, the reactor container can be cooled by water in the suppression chamber by a static means without requiring pumps. (I.S.)

  4. Container Closure Integrity Testing of Prefilled Syringes.

    Science.gov (United States)

    Peláez, Sarah S; Mahler, Hanns-Christian; Matter, Anja; Koulov, Atanas; Singh, Satish K; Germershaus, Oliver; Mathaes, Roman

    2018-04-04

    Prefilled syringes (PFSs) are increasingly preferred over vials as container closure systems (CCSs) for injectable drug products when facilitated or self-administration is required. However, PFSs are more complex compared to CCSs consisting of vial, rubber stopper and crimp cap. Container closure integrity (CCI) assurance and verification has been a specific challenge for PFSs as they feature several sealing areas. A comprehensive understanding of the CCS is necessary for an appropriate CCI assessment as well as for packaging development and qualification. A comprehensive CCI assessment of six different PFSs from three different manufacturers (including one polymeric PFS) was conducted using helium leak testing. PFS components were manipulated to systematically assess the contribution of the different sealing areas to CCI, namely rigid needle shield (RNS)/needle, RNS/tip cone and the individual ribs of a syringe plunger. The polymeric PFS required an equilibrium measurement for accurate CCIT. The different sealing areas and a single plunger rib were shown to provide adequate CCI. Acceptable tip cap movement until the point of CCI failure was estimated. The assessment of acceptable tip cap movement demonstrated the importance of considering the RNS/tip cone seal design to ensure CCI of the PFS upon post assembly possesses and shipment. Copyright © 2018. Published by Elsevier Inc.

  5. Containment design, performance criteria and research needs for advanced reactor designs

    International Nuclear Information System (INIS)

    Bagdi, G.; Ali, S.; Costello, J

    2004-01-01

    This paper points out some important shifts in the basic expectations in the performance requirements for containment structures and discusses the areas where the containment structure design requirements and acceptance criteria can be integrated with ultimate test based insights. Although there has not been any new reactor construction in the United States for over thirty years, several designs of evolutionary and advanced reactors have already been certified. Performance requirements for containment structures under design basis and severe accident conditions and explicit consideration of seismic margins have been used in the design certification process. In the United States, the containment structure design code is the American Society of Mechanical Engineers, Boiler and Pressure Vessel Code, Section III, Division 1, Subsection NE-Class MC for the steel containment and Section III, Division 2 for reinforced and prestressed concrete reactor vessels and containments. This containment design code was based on the early concept of applying design basis internal pressure and associated load combinations that included the operating basis and safe shutdown earthquake ground motion. These early design criteria served the nuclear industry and the regulatory authorities in maintaining public health and safety. However, these early design criteria do not incorporate the performance criteria related to containment function in an integrated fashion. Research in large scale model testing of containment structures to failure from over pressurization and shake table testing using simulated ground motion, have produced insights related to failure modes and material behavior at failure. The results of this research provide the opportunity to integrate these observations into design and acceptance criteria. This integration process would identify 'gaps' in the present knowledge and future research needs. This knowledge base is important for gleaning risk-informed insights into

  6. Investigation of the transportation requirements for fusion power plants

    International Nuclear Information System (INIS)

    Rhoads, R.E.; Davis, D.K.

    1976-09-01

    This report presents a general investigation of the transport requirements associated with the construction and operation of conceptual fusion reactors. Projections of amounts of construction and operating materials requiring transportation are presented for several proposed designs. The material to be shipped is described along with the shipping containers that might be used, the transport modes and the expected impact of transporting these materials. Transportation of both radioactive and nonradioactive materials will be required. Most of these materials are routinely shipped by the transportation industry. Transportation requirements of a representative fusion reactor are also compared with Liquid Metal Fast Breeder Reactor (LMFBR) requirements

  7. Exploratory double-blind, parallel-group, placebo-controlled study of edaravone (MCI-186) in amyotrophic lateral sclerosis (Japan ALS severity classification: Grade 3, requiring assistance for eating, excretion or ambulation).

    Science.gov (United States)

    2017-10-01

    Our objective was to explore the efficacy and safety of edaravone in amyotrophic lateral sclerosis (ALS) patients with a Japan ALS severity classification of Grade 3. In a 24-week, double-blind, randomized study, 25 patients who met all of the following criteria were enrolled: Japan ALS severity classification Grade 3; definite, probable, or probable-laboratory supported ALS (El Escorial/revised Airlie House); forced vital capacity (%FVC) ≥60%; duration of disease ≤3 years at consent; and change in the revised ALS functional rating scale (ALSFRS-R) score of -1 to -4 points during the 12-week pre-observation period. Patients received edaravone (n = 13) or placebo (n = 12) for six cycles. The efficacy outcome was change in the ALSFRS-R score. The least-squares mean change in the ALSFRS-R score ± standard error during the 24-week treatment was -6.52 ± 1.78 in the edaravone group and -6.00 ± 1.83 in the placebo group; the difference of -0.52 ± 2.46 was not statistically significant (p = 0.835). Incidence of adverse events was 92.3% (12/13) in the edaravone group and 100.0% (12/12) in the placebo group. There was no intergroup difference in the changes in the ALSFRS-R score. The incidences of adverse events were similar in the two groups.

  8. ACR-1000: Enhanced response to severe accidents

    International Nuclear Information System (INIS)

    Popov, N.K.; Santamaura, P.; Shapiro, H.; Snell, V.G.

    2006-01-01

    Full text: Atomic Energy of Canada Limited (AECL) developed the Advanced CANDU Reactor-TM700 (ACR-700TM) as an evolutionary advancement of the current CANDU 6R reactor. As further advancement of the ACR design, AECL is currently developing the ACR-1000TM for the Canadian and international market. The ACR-1000 is aimed at producing electrical power for a capital cost and a unit-energy cost significantly less than that of the current generation of operating nuclear plants, while achieving shorter construction schedule, high plant capacity factor, improved operations and maintenance, increased operating life. and enhanced safety features. The reference ACR-1000 plant design is based on an integrated two-unit plant, using enriched fuel and light-water coolant, with each unit having a nominal gross output of about 1200 MWe. This paper presents the ACR-1000 features that are designed to mitigate limited core damage and severe core damage states, including core retention within vessel, core damage termination, and containment integrity maintenance. Core retention within vessel in CANDU-type reactors includes both retention within fuel channels, and retention within the calandria vessel. The moderator heavy water in the ACR-1000 calandria vessel, as in any other CANDU-type reactor, provides ample heat removal capacity in severe accidents. The ACR-1000 calandria vessel design permits for passive rejection of decay heat from the moderator to the shield water. Also, the calandria vessel will be designed for debris retention. Core damage termination is achieved by flooding of the core components with water and keeping them flooded thereafter. Successful termination can be achieved in the fuel channels, calandria vessel or calandria vault by water supply by the Long Term Cooling (LTC) pumps and by gravity feed from the Reserve Water System. The ACR-1000 containment is required to withstand external events such as earthquakes, tornados, floods and aircraft crashes. Containment

  9. Steps required to inclusion in commercial ECG analysis systems--the new ECG indices for quantitating extent, acuteness and severity of acute myocardial ischemia for facilitating emergency triage decisions.

    Science.gov (United States)

    Hampton, David R

    2014-01-01

    Clinically useful diagnostic methods for chest pain triage often fail to reach everyday practice where they can improve patient outcomes. One means to bridge the gap is through adoption of ECG interpretive algorithms with enhanced accuracy or expanded features into established commercial products. The transition from innovation to industry can be facilitated if researchers consider three factors aiding a successful handoff to companies. First, they should assess their algorithm to assure that it meets a real market need and can be easily assimilated by commercial partners. Second, their design documentation and databases should support the regulated development processes required of manufacturers. Finally, they should hold appropriate expectations for the structure of commercial partnerships that lead to release of a marketed product. Copyright © 2014 Elsevier Inc. All rights reserved.

  10. Power Burst Facility Severe Fuel Damage test series

    International Nuclear Information System (INIS)

    Buescher, B.J.; Osetek, D.J.; Ploger, S.A.

    1982-01-01

    The Severe Fuel Damage (SFD) tests planned for the Power Burst Facility (PBF) are described. Bundles containing 32 zircaloy-clad, PWR-type fuel rods will be subjected to severe overheating transients in a high-pressure, superheated-steam environment. Cladding temperatures are expected to reach 2400 0 K, resulting in cladding ballooning and rupture, severe cladding oxidation, cladding melting, fuel dissolution, fuel rod fragmentation, and possibly, rubble bed formation. An experiment effluent collection system is being installed and the PBF fission product monitoring system is being upgraded to meet the special requirements of the SFD tests. Scoping calculations were performed to evaluate performance of the SFD test design and to establish operational requirements for the PBF loop

  11. Nuclear containment systems and in-service inspection status of Korea nuclear power plants

    Energy Technology Data Exchange (ETDEWEB)

    Jihong, Park; Jaekeun, Hong; Banuk, Park [Korea Institute of Machinery and Materials, Dept. of Authorized Test and Evaluation, Kyungnam (Korea, Republic of)

    2007-07-01

    20 unit nuclear power plants in Korea have been operated and maintained since the first unit started in commercial service in 1978. Most recently 4 units were under construction and several units were planned to be constructed. by industries. 4 types of nuclear containment systems have been constructed until now: first, metal containments, then pre-stressed concrete containments with grouted tendon systems, followed by pre-stressed concrete containments with un-grouted tendon systems, and Korea standard nuclear containments. All the nuclear containments should be inspected periodically. Therefore for periodic in-service inspection, several appropriate technical requirements should be applied differently depending on the specific nuclear containment types. With the changes of times, nuclear containment systems have undergone a remarkable change, and finally nuclear containment system of Korea standard nuclear power plant was settled down, and as a matter of course it dominates the trend of present and future nuclear containment systems. Overall in-service inspection results of most Korea nuclear containments have not showed any serious evidence of degradation.

  12. Ahp2 (Hop2) function in Arabidopsis thaliana (Ler) is required for stabilization of close alignment and synaptonemal complex formation except for the two short arms that contain nucleolus organizer regions.

    Science.gov (United States)

    Stronghill, P; Pathan, N; Ha, H; Supijono, E; Hasenkampf, C

    2010-08-01

    A cytological comparative analysis of male meiocytes was performed for Arabidopsis wild type and the ahp2 (hop2) mutant with emphasis on ahp2's largely uncharacterized prophase I. Leptotene progression appeared normal in ahp2 meiocytes; chromosomes exhibited regular axis formation and assumed a typical polarized nuclear organization. In contrast, 4',6'-diamidino-2-phenylindole-stained ahp2 pachytene chromosome spreads demonstrated a severe reduction in stabilized pairing. However, transmission electron microscopy (TEM) analysis of sections from meiocytes revealed that ahp2 chromosome axes underwent significant amounts of close alignment (44% of total axis). This apparent paradox strongly suggests that the Ahp2 protein is involved in the stabilization of homologous chromosome close alignment. Fluorescent in situ hybridization in combination with Zyp1 immunostaining revealed that ahp2 mutants undergo homologous synapsis of the nucleolus-organizer-region-bearing short arms of chromosomes 2 and 4, despite the otherwise "nucleus-wide" lack of stabilized pairing. The duration of ahp2 zygotene was significantly prolonged and is most likely due to difficulties in chromosome alignment stabilization and subsequent synaptonemal complex formation. Ahp2 and Mnd1 proteins have previously been shown, "in vitro," to form a heterodimer. Here we show, "in situ," that the Ahp2 and Mnd1 proteins are synchronous in their appearance and disappearance from meiotic chromosomes. Both the Ahp2 and Mnd1 proteins localize along the chromosomal axis. However, localization of the Ahp2 protein was entirely foci-based whereas Mnd1 protein exhibited an immunostaining pattern with some foci along the axis and a diffuse staining for the rest of the chromosome.

  13. Study of the ruthenium fission-product behavior in the containment, in the case of a nuclear reactor severe accident; Etude du comportement du produit de fission ruthenium dans l'enceinte de confinement d'un reacteur nucleaire, en cas d'accident grave

    Energy Technology Data Exchange (ETDEWEB)

    Mun, Ch

    2007-03-15

    Ruthenium tetroxide is an extremely volatile and highly radio-toxic species. During a severe accident with air ingress in the reactor vessel, ruthenium oxides may reach the reactor containment building in significant quantities. Therefore, a better understanding of the RuO{sub 4}(g) behaviour in the containment atmosphere is of primary importance for the assessment of radiological consequences, in the case of potential releases of this species into the environment. A RuO{sub 4}(g) decomposition kinetic law was determined. Steam seems to play a catalytic role, as well as the presence of ruthenium dioxide deposits. The temperature is also a key parameter. The nature of the substrate, stainless steel or paint, did not exhibit any chemical affinities with RuO{sub 4}(g). This absence of reactivity was confirmed by XPS analyses, which indicate the presence of the same species in the Ru deposits surface layer whatever the substrates considered. It has been concluded that RuO{sub 4}(g) decomposition corresponds to a bulk gas phase decomposition. The ruthenium re-volatilization phenomenon under irradiation from Ru deposits was also highlighted. An oxidation kinetic law was determined. The increase of the temperature and the steam concentration promote significantly the oxidation reaction. The establishment of Ru behavioural laws allowed making a modelling of the Ru source term. The results of the reactor calculations indicate that the values obtained for {sup 106}Ru source term are closed to the reference value considered currently by the IRSN, for 900 MWe PWR safety analysis. (author)

  14. On severe accident hydrogen behaviour in Loviisa

    International Nuclear Information System (INIS)

    Okkonen, T.

    1996-02-01

    This study is related to the hydrogen management strategy of the Loviisa ice-condenser containments. A synthetic survey is conducted of the various parts of the subject by using compact 'back-of-the-envelope' analysis methods. The analysed cases are consistent with the principal hydrogen management approaches proposed by the utility Imatran Voima Oy (IVO). The study begins by introduction of the Loviisa plant features and various severe accident types. Hydrogen generation characteristics are analysed mainly for the core degradation phase, but the hydrogen sources from molten fuel-coolant interactions and reflooding of a degraded core are discussed, as well. The hydrogen generation and release rates are compared with the overall gas convection and mixing conditions in order to estimate hydrogen concentrations in the containment. The natural convection currents are examined also from the scaling point of view, concerning the scaled-down VICTORIA tests of IVO. Finally, the potential for large deflagration loadings or local detonations is examined for the Loviisa containments. The study is concluded by preliminary subjective judgments about the most critical factors of the Loviisa hydrogen problematics and about any issues that may require additional confirmative research. (orig.) (47 refs., 4 figs., 24 tabs.)

  15. On severe accident hydrogen behaviour in Loviisa

    Energy Technology Data Exchange (ETDEWEB)

    Okkonen, T. [OTO-Consulting Ay, Helsinki (Finland)

    1996-02-01

    This study is related to the hydrogen management strategy of the Loviisa ice-condenser containments. A synthetic survey is conducted of the various parts of the subject by using compact `back-of-the-envelope` analysis methods. The analysed cases are consistent with the principal hydrogen management approaches proposed by the utility Imatran Voima Oy (IVO). The study begins by introduction of the Loviisa plant features and various severe accident types. Hydrogen generation characteristics are analysed mainly for the core degradation phase, but the hydrogen sources from molten fuel-coolant interactions and reflooding of a degraded core are discussed, as well. The hydrogen generation and release rates are compared with the overall gas convection and mixing conditions in order to estimate hydrogen concentrations in the containment. The natural convection currents are examined also from the scaling point of view, concerning the scaled-down VICTORIA tests of IVO. Finally, the potential for large deflagration loadings or local detonations is examined for the Loviisa containments. The study is concluded by preliminary subjective judgments about the most critical factors of the Loviisa hydrogen problematics and about any issues that may require additional confirmative research. (orig.) (47 refs., 4 figs., 24 tabs.).

  16. Investigations on passive containment cooling

    International Nuclear Information System (INIS)

    Knebel, J.U.; Cheng, X.; Neitzel, H.J.; Erbacher, F.J.; Hofmann, F.

    1997-01-01

    The composite containment design for advanced LWRs that has been examined under the PASCO project is a promising design concept for purely passive decay heat removal after a severe accident. The passive cooling processes applied are natural convection and radiative heat transfer. Heat transfer through the latter process removes at an emission coefficient of 0.9 about 50% of the total heat removed via the steel containment, and thus is an essential factor. The heat transferring surfaces must have a high emission coefficient. The sump cooling concept examined under the SUCO project achieves a steady, natural convection-driven flow from the heat source to the heat sink. (orig./CB) [de

  17. A performance-oriented and risk-based regulation for containment testing

    International Nuclear Information System (INIS)

    Dey, M.

    1994-01-01

    In August 1992, the NRC initiated a major initiative to develop requirements for containment testing that are less prescriptive, and more performance-oriented and risk-based. This action was a result of public comments and several studies that concluded that the economic burden of certain, present containment testing requirements are not commensurate with their safety benefits. The rulemaking will include consideration of relaxing the allowable containment leakage rate, increasing the interval for the integrated containment test, and establishing intervals for the local containment leak rate tests based on their performance. A study has been conducted to provide technical information for establishing the performance criteria for containment tests, the allowable leakage rate, commensurate with its significance to total public risk. The study used results of a recent comprehensive study conducted by the NRC, NUREG-1150, 'Severe Accident Risks: An Assessment for Five U.S. Nuclear Power Plants,' to examine the sensitivity of containment leakage to public risk. Risk was found to be insensitive to containment leakage rate up to levels of about 100 percent-volume per day for certain types of containments. PRA methods have also been developed to establish risk-based intervals for containment tests based on their past experience. Preliminary evaluations show that increasing the interval for the integrated containment leakage test from three times to once every ten years would have an insignificant impact on public risk. Preliminary analyses of operational experience data for local leak rate tests show that performance-based testing, valves and penetrations that perform well are tested less frequently, is feasible with marginal impact on safety. The above technical studies are being used to develop efficient (cost-effective) requirements for containment tests. (author). 4 refs., 2 figs

  18. Periodic inspection of CANDU nuclear power plant containment components

    International Nuclear Information System (INIS)

    1989-09-01

    This Standard is one in a series intended to provide uniform requirements for CANDU nuclear power plants. It provides requirements for the periodic inspection of containment components including the containment pressure suppression systems

  19. Biologic assessment of copper-containing amalgams.

    Science.gov (United States)

    Mjor, I A; Eriksen, H M; Haugen, E; Skogedal, O

    1977-12-01

    In order to reduce creep and avoid marginal fractures in amalgam restorations, new alloys containing higher proportions of copper have been introduced. Fillings of these materials were placed in cavities prepared in the deciduous teeth of monkeys or placed in polyethylene tubes and implanted subcutaneously in rats. Conventional silver/tin alloys and zinc oxide eugenol cement were used as reference materials. Despite limitations due to the varying depths of cavities and the small number of animals involved it was concluded that the high copper alloys caused more severe pulp damage than the other materials studied. In the implantation studies many of the high copper specimens were exfoliated before the end of the experimental period. It is concluded that in deep cavities these materials require the use of a non-toxic base or lining material although as they are commonly used in young children's teeth the placement of linings and the isolation of the cavity pose problems.

  20. 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.)

  1. Do pyrotechnics contain radium?

    Energy Technology Data Exchange (ETDEWEB)

    Steinhauser, Georg; Musilek, Andreas, E-mail: georg.steinhauser@ati.ac.a [Vienna University of Technology, Atominstitut der Oesterreichischen Universitaeten, Stadionallee 2, A-1020 Wien (Austria)

    2009-07-15

    Many pyrotechnic devices contain barium nitrate which is used as an oxidizer and colouring agent primarily for green-coloured fireworks. Similarly, strontium nitrate is used for red-coloured pyrotechnic effects. Due to their chemical similarities to radium, barium and strontium ores can accumulate radium, causing a remarkable activity in these minerals. Radium in such contaminated raw materials can be processed together with the barium or strontium, unless extensive purification of the ores was undertaken. For example, the utilization of 'radiobarite' for the production of pyrotechnic ingredients can therefore cause atmospheric pollution with radium aerosols when the firework is displayed, resulting in negative health effects upon inhalation of these aerosols. In this study, we investigated the occurrence of gamma-photon-emitting radionuclides in several pyrotechnic devices. The highest specific activities were due to K-40 (up to 20 Bq g{sup -1}, average value 14 Bq g{sup -1}). Radium-226 activities were in the range of 16-260 mBq g{sup -1} (average value 81 mBq g{sup -1}). Since no uranium was found in any of the samples, indeed, a slight enrichment of Ra-226 in coloured pyrotechnics can be observed. Radioactive impurities stemming from the Th-232 decay chain were found in many samples as well. In the course of novel developments aiming at the 'greening' of pyrotechnics, the potential radioactive hazard should be considered as well.

  2. Do pyrotechnics contain radium?

    International Nuclear Information System (INIS)

    Steinhauser, Georg; Musilek, Andreas

    2009-01-01

    Many pyrotechnic devices contain barium nitrate which is used as an oxidizer and colouring agent primarily for green-coloured fireworks. Similarly, strontium nitrate is used for red-coloured pyrotechnic effects. Due to their chemical similarities to radium, barium and strontium ores can accumulate radium, causing a remarkable activity in these minerals. Radium in such contaminated raw materials can be processed together with the barium or strontium, unless extensive purification of the ores was undertaken. For example, the utilization of 'radiobarite' for the production of pyrotechnic ingredients can therefore cause atmospheric pollution with radium aerosols when the firework is displayed, resulting in negative health effects upon inhalation of these aerosols. In this study, we investigated the occurrence of gamma-photon-emitting radionuclides in several pyrotechnic devices. The highest specific activities were due to K-40 (up to 20 Bq g -1 , average value 14 Bq g -1 ). Radium-226 activities were in the range of 16-260 mBq g -1 (average value 81 mBq g -1 ). Since no uranium was found in any of the samples, indeed, a slight enrichment of Ra-226 in coloured pyrotechnics can be observed. Radioactive impurities stemming from the Th-232 decay chain were found in many samples as well. In the course of novel developments aiming at the 'greening' of pyrotechnics, the potential radioactive hazard should be considered as well.

  3. Hydrogen distribution analysis for CANDU 6 containment using the GOTHIC containment analysis code

    International Nuclear Information System (INIS)

    Nguyen, T.H.; Collins, W.M.

    1995-01-01

    Hydrogen may be generated in the reactor core by the zircaloy-steam reaction for a postulated loss of coolant accident (LOCA) scenario with loss of emergency core cooling (ECC). It is important to predict hydrogen distribution within containment in order to determine if flammable mixtures exist. This information is required to determine the best locations in containment for the placement of mitigation devices such as igniters and recombiners. For large break loss coolant accidents, hydrogen is released after the break flow has subsided. Following this period of high discharge the flow in the containment building undergoes transition from forced flow to a buoyancy driven flow (particularly when local air coolers (LACS) are not credited). One-dimensional computer codes (lumped parameter) are applicable during the initial period when a high degree of mixing occurs due to the forced flow generated by the break. However, during the post-blowdown phase the assumption of homogeneity becomes less accurate, and it is necessary to employ three-dimensional codes to capture local effects. This is particularly important for purely buoyant flows which may exhibit stratification effects. In the present analysis a three-dimensional model of CANDU 6 containment was constructed with the GOTHIC computer code using a relatively coarse mesh adequate enough to capture the salient features of the flow during the blowdown and hydrogen release periods. A 3D grid representation was employed for that portion of containment in which the primary flow (LOCA and post-LOCA) was deemed to occur. The remainder of containment was represented by lumped nodes. The results of the analysis indicate that flammable concentrations exist for several minutes in the vicinity of the break and in the steam generator enclosure. This is due to the fact that the hydrogen released from the break is primarily directed upwards into the steam generator enclosure due to buoyancy effects. Once hydrogen production ends

  4. Whole exome sequencing identified 1 base pair novel deletion in BCL2-associated athanogene 3 (BAG3) gene associated with severe dilated cardiomyopathy (DCM) requiring heart transplant in multiple family members.

    Science.gov (United States)

    Rafiq, Muhammad Arshad; Chaudhry, Ayeshah; Care, Melanie; Spears, Danna A; Morel, Chantal F; Hamilton, Robert M

    2017-03-01

    Dilated cardiomyopathy (DCM) is characterized by dilation and impaired contraction of the left ventricle or both ventricles. Among hereditary DCM, the genetic causes are heterogeneous, and include mutations encoding cytoskeletal, nucleoskeletal, mitochondrial, and calcium-handling proteins. We report three severely affected males, in a four-generation pedigree, with DCM phenotype who underwent cardiac transplant. Cardiomegaly with marked biventricular dilation and fibrosis were noticeable histopathological findings. The affected males had tested negative on a 46-gene pancardiomyopathy panel. Whole Exome Sequencing (WES) was performed to reveal mutation in the gene responsible in generation of DCM phenotypes. The 1-bp (Chr10:121435979delC; c.913delC) novel heterozygous deletion in exon 4 of BAG3, was identified in three affected males, resulted in frame-shift and a premature termination codon (p.Met306-Stop) producing a truncated BAG3 protein lacking functionally important PXXP and BAG domains. WES data were further utilized to map 10 SNP markers around the discovered mutation to generate shared disease haplotype in all affected individuals encompassing 11 Mb on 10q25.3-26.2 harboring BAG3. Finally genotypes were inferred for the unavailable/deceased individuals in the pedigrees. Here we propose that Chr10:121435979delC in BAG3 is a causal mutation in these subjects. Our and earlier studies indicate that BAG3 mutations are associated with DCM phenotypes. BAG3 should be added to cardiomyopathy gene panels for screening of DCM patients, and patients previously considered gene elusive should undergo sequencing of the BAG3 gene. © 2017 Wiley Periodicals, Inc. © 2017 Wiley Periodicals, Inc.

  5. Partiality and Container Monads

    DEFF Research Database (Denmark)

    Uustalu, Tarmo; Veltri, Niccolò

    2017-01-01

    the relationship between containers and lifting monads. We show that the lifting monads usually employed in type theory can be specified in terms of containers. Moreover, we give a precise characterization of containers whose interpretations carry a lifting monad structure. We show that these conditions...

  6. Ultimate internal pressure capacity of concrete containment structures

    International Nuclear Information System (INIS)

    Krishnaswamy, C.N.; Namperumal, R.; Al-Dabbagh, A.

    1983-01-01

    Lesson learned from the accident at Three-Mile Island nuclear plant has necessitated the computation of the ultimate internal pressure capacity of containment structures as a licensing requirement in the U.S. In general, a containment structure is designed to be essentially elastic under design accident pressure. However, as the containment pressure builds up beyond the design value due to a more severe postulated accident, the containment response turns nonlinear as it sequentially passes through cracking of concrete, yielding of linear plate, yielding of rebar, and yielding of post-tensioning tendon (if the containment concrete is prestressed). This paper reports on the determination of the ultimate internal pressure capacity and nonlinear behavior of typical reinforced and prestressed concrete BWR containments. The probable modes of failure, the criteria for ultimate pressure capacity, and the most critical sections are described. Simple equations to hand-calculate the ultimate pressure capacity and the nonlinear behavior at membrane sections of the containment shell are presented. A nonlinear finite element analysis performed to determine the nonlinear behavior of the entire shell including nonmembrane sections is briefly discribed. The analysis model consisted of laminated axisymmetric shell finite elements with nonlinear stress-strain properties for each material. Results presented for typical BWR concrete containments include nonlinear response plots of internal pressure versus containment deflection and strains in the liner, rebar, and post-tensioning tendons at the most stressed section in the shell. Leak-tightness of the containment liner and the effect of thermal loads on the ultimate capacity are discussed. (orig.)

  7. 1988-year of high integrity container evaluation, controversy and regulatory action

    International Nuclear Information System (INIS)

    Jones, D.

    1989-01-01

    During 1988, the Nuclear Regulatory Commission(NRC) completed review of and prepared technical evaluation reports on several topical reports describing containers designed to meet the 10CFR61 requirements for high integrity containers (HICs). An all metal Ferralium container and a stainless steel/polyethylene lined container were approved by the NRC. However, the NRC did not approve any containers designed from polyethylene material. The NRC staff concluded that polyethylene containers do not meet the structural stability requirements of Part 61, and unless they are combined with some engineered structure or overpack, they are not adequate for disposal of low-level radioactive wastes that require disposal in a structurally stable form. In conflict with these NRC findings, the State of South Carolina Department of Health and Environmental Control (SCDHEC) has given interim approval for continued use of polyethylene containers at the Barnwell disposal site with some restrictions on how the containers are buried depending on the classification of the waste. This paper reviews the applicable federal regulations, presents a chronology of events describing how the controversy over high integrity containers evolved from 1980 to 1989, summarizes the technical issues involved and suggests an approach that waste generators should follow during this situation of regulatory uncertainty

  8. Stowing the Right Containers on Container Vessels

    DEFF Research Database (Denmark)

    Jensen, Rune Møller

    2014-01-01

    ’s largest container vessels using standard mathematical programming techniques and off-the-shelf solvers. The presentation will provide basic insight into the domain, with pointers to further information that enable you to join in this promising new path of operations research and business....

  9. Discussion of important safety requirements for new nuclear power plants

    International Nuclear Information System (INIS)

    Zhang Lin; Jia Xiang; Yan Tianwen; Li Wenhong; Li Chun

    2014-01-01

    This paper presents the analysis of several important safety requirements and improvement direction. Technical view of security goals on site safety evaluation, internal and external events fortification, serious accident prevention and mitigation, as well as the core, containment system and instrument control system design and engineering optimization, and etc are indicated. It will be useful for new plant design, construction and safety improvement. (authors)

  10. Several crimes solved

    CERN Multimedia

    Relations with the Host States Service

    2007-01-01

    A member of a contractor's personnel suspected of having committed several thefts in and around Building 180 has recently been questioned by the French police. He was immediately tried by the court in Bourg-en-Bresse and sentenced to six months in prison, with a requirement to serve at least three months. His arrest was facilitated, among other things, by a video recording, fast and detailed statements to the CERN Fire Brigade and close collaboration between the members of the personnel concerned, the Reception and Access Control Service and the police. Several laptops and other items of electronic equipment were seized during a search of the culprit's home. A stolen digital camera has yet to be returned to its owner as he has not reported the theft to the CERN Fire Brigade and the police. The person concerned is therefore requested to go to the Gendarmerie in Saint-Genis-Pouilly with the necessary proof of ownership. In addition, the French authorities have informed CERN that the presumed authors of the a...

  11. Tank waste remediation system functions and requirements document

    International Nuclear Information System (INIS)

    Carpenter, K.E

    1996-01-01

    This is the Tank Waste Remediation System (TWRS) Functions and Requirements Document derived from the TWRS Technical Baseline. The document consists of several text sections that provide the purpose, scope, background information, and an explanation of how this document assists the application of Systems Engineering to the TWRS. The primary functions identified in the TWRS Functions and Requirements Document are identified in Figure 4.1 (Section 4.0) Currently, this document is part of the overall effort to develop the TWRS Functional Requirements Baseline, and contains the functions and requirements needed to properly define the top three TWRS function levels. TWRS Technical Baseline information (RDD-100 database) included in the appendices of the attached document contain the TWRS functions, requirements, and architecture necessary to define the TWRS Functional Requirements Baseline. Document organization and user directions are provided in the introductory text. This document will continue to be modified during the TWRS life-cycle

  12. Tank waste remediation system functions and requirements document

    Energy Technology Data Exchange (ETDEWEB)

    Carpenter, K.E

    1996-10-03

    This is the Tank Waste Remediation System (TWRS) Functions and Requirements Document derived from the TWRS Technical Baseline. The document consists of several text sections that provide the purpose, scope, background information, and an explanation of how this document assists the application of Systems Engineering to the TWRS. The primary functions identified in the TWRS Functions and Requirements Document are identified in Figure 4.1 (Section 4.0) Currently, this document is part of the overall effort to develop the TWRS Functional Requirements Baseline, and contains the functions and requirements needed to properly define the top three TWRS function levels. TWRS Technical Baseline information (RDD-100 database) included in the appendices of the attached document contain the TWRS functions, requirements, and architecture necessary to define the TWRS Functional Requirements Baseline. Document organization and user directions are provided in the introductory text. This document will continue to be modified during the TWRS life-cycle.

  13. Passive containment system

    International Nuclear Information System (INIS)

    Kleimola, F.W.

    1977-01-01

    Disclosed is a containment system that provides complete protection entirely by passive means for the loss of coolant accident in a nuclear power plant and wherein all stored energy released in the coolant blowdown is contained and absorbed while the nuclear fuel is prevented from over-heating by a high containment back-pressure and a reactor vessel refill system. The primary containment vessel is restored to a high sub-atmospheric pressure within a few minutes after accident initiation and the decay heat is safely transferred to the environment while radiolytic hydrogen is contained by passive means. 20 claims, 14 figures

  14. Control room habitability during severe accidents

    International Nuclear Information System (INIS)

    Siu, R.P.

    1989-01-01

    The requirements for protection of control room personnel against radiation hazards are specified in 10CFR50, Appendix A, GDC 19. The conventional approach involves a mechanistic evaluation of the radiation doses to control room personnel during design-basis accidents. In this study, an assessment of control room habitability during severe accidents is conducted. The potential levels of radiation hazards to control room personnel are evaluated in terms of both magnitude and probability of occurrence. The expected values for the probabilities of exceeding GDC-19 limits and the cumulative probability distributions of control room doses are determined. In this study, a pressurized water reactor with a large dry containment has been selected for analysis. The types of control rooms evaluated in this study include designs with: (a) filtered local intakes only, (b) filtered recirculation only, (c) filtered local intakes and recirculation, and (d) filtered dual remote intakes and recirculation. From the observations, it is concluded that, except for control room D, all other control room designs may require improvements in order to provide adequate radiation protection during severe accidents, particularly in terms of reducing whole-body gamma doses and skin doses. Potential design improvements include reduction of intake flows for concepts relying on pressurization, reduction in overall leakages, and control room pressurization through the use of bottled air supply

  15. Severe neurotoxicity following ingestion of tetraethyl lead.

    Science.gov (United States)

    Wills, Brandon K; Christensen, Jason; Mazzoncini, Joe; Miller, Michael

    2010-03-01

    Organic lead compounds are potent neurotoxins which can result in death even from small exposures. Traditionally, these compounds are found in fuel stabilizers, anti-knock agents, and leaded gasoline. Cases of acute organic lead intoxication have not been reported for several decades. We report a case of a 13-year-old Iraqi male who unintentionally ingested a fuel stabilizer containing 80-90% tetraethyl lead, managed at our combat support hospital. The patient developed severe neurologic symptoms including agitation, hallucinations, weakness, and tremor. These symptoms were refractory to escalating doses of benzodiazepines and ultimately required endotracheal intubation and a propofol infusion. Adjunctive therapies included chelation, baclofen, and nutrition provided through a gastrostomy tube. The patient slowly recovered and was discharged in a wheelchair 20 days after ingestion, still requiring tube feeding. Follow-up at 62 days post-ingestion revealed near-resolution of symptoms with residual slurred speech and slight limp. This case highlights the profound neurotoxic manifestations of acute organic lead compounds.

  16. Emergency air cleaning system development for LMFBR containments

    International Nuclear Information System (INIS)

    McCormack, J.D.; Hilliard, R.K.; Postma, A.K.; Muhlestein, L.D.

    1975-01-01

    Criteria for evaluating the various types of Emergency Air Cleaning Systems which may be used in LMFBR plants have been established for both single containment and containment-confinement arrangements. These two plant arrangements have quite different air cleaning requirements for postulated design base accident conditions. Work is currently in progress to select from a list of candidate air cleaning systems those which best meet the criteria requirements. By means of a weighted rating system, areas of strength or weakness can be found and the conceptual system design then optimized. The final system arrangements will be ranked and several of the most promising systems selected for large-scale tests in the former CSE vessel at Hanford. 8 references. (U.S.)

  17. Fibre-concrete container

    International Nuclear Information System (INIS)

    2000-01-01

    In this leaflet the fibre-concrete container for radioactive wastes is described. The fibre container is made of fibre-concrete that contains cement, aggregate, sand, filter, flame-silica, super-plastificator, water and scattered metal fibres. The fibre-concrete container has a dice shape with outer dimension 1.7 x 1.7 x 1.7 m. It is mounted of a container body, a container cover and two caps. Total weight of container is 4,240 kg, maximum weight of loaded container do not must exceed 15,000 kg. The physical and mechanical properties of the fibre-concrete container are described in detail. The fibre-concrete container manufactured for storing of low and intermediate radioactive wastes. A fibre-concrete container utilization to store of radioactive wastes solves these problems: increase of stability of stored packages of radioactive waste; watertightness within 300 years at least; static stability of bearing space; better utilization of bearing spaces; insulation of radioactive waste in a case of seismic and geological event; increase of fire resistance; and transport of radioactive waste

  18. Containment integrity research program plan

    International Nuclear Information System (INIS)

    1987-08-01

    This report presents a plan for research on the question of containment performance in postulated severe accident scenarios. It focuses on the research being performed by the Structural and Seismic Engineering Branch, Division of Engineering, Office of Nuclear Regulatory Research. Summaries of the plans for this work have previously been published in the ''Nuclear Power Plant Severe Accident Research Plan'' (NUREG-0900). This report provides an update to reflect current status. This plan provides a summary of results to date as well as an outline of planned activities and milestones to the contemplated completion of the program in FY 1989

  19. Shipment and Storage Containers for Tritium Production Transportation Casks

    International Nuclear Information System (INIS)

    Massey, W.M.

    1998-04-01

    The need for a shipping and storage container for the Tritium production transportation casks is addressed in this report. It is concluded that a shipping and storage container is not required. A recommendation is made to eliminate the requirement for this container because structural support and inerting requirements can be satisfied completely by the cask with a removable basket

  20. Plutonium accident resistant container project

    International Nuclear Information System (INIS)

    Andersen, J.A.

    1978-09-01

    The PARC (plutonium accident resistant container) project resulted in the design, development, and certification testing of a crashworthy air-transportable plutonium package (shipping container) for certification by the USNRC (Nuclear Regulatory Commission). This PAT-1 (plutonium air transportable) package survives a very severe sequential test program of impact, crush, puncture, slash, burn, and water immersion. There is also an individual hydrostatic pressure test. The package has a payload mass capacity of 2 kg of PuO 2 and a thermal capacity of 25 watts. The design rationale for very high energy absorption (impact, crush, puncture, and slash protection) with residual high-level fire protection, resulted in a reasonably small air-transportable package, advancing the packaging state-of-art. Optimization design iterations were utilized in the areas of impact energy absorption and stress and thermal analysis. Package test results are presented in relation to radioactive materials containment acceptance criteria, shielding and criticality standards

  1. Plutonium accident resistant container project

    International Nuclear Information System (INIS)

    Andersen, J.A.

    1978-05-01

    The PARC (plutonium accident resistant container) project resulted in the design, development, and certification testing of a crashworthy air-transportable plutonium package (shipping container) for certification by the USNRC. This PAT-1 (plutonium air transportable) package survives a very severe sequential test program of impact, crush, puncture, slash, burn, and water immersion. There is also an individual hydrostatic pressure test. The package has a payload mass capacity of 2 kg of PuO2 and a thermal capacity of 25 watts. The design rationale for very high energy absorption (impact, crush, puncture, and slash protection) with residual high-level fire protection, resulted in a reasonalby small air-transportable package, advancing the packaging state-of-art. Optimization design iterations were utilized in the areas of impact energy absorption and stress and thermal analysis. Package test results are presented in relation to radioactive materials containment acceptance criteria, shielding and criticality standards

  2. TRUPACT-II Container Maintenance Program Plan

    International Nuclear Information System (INIS)

    1991-05-01

    This document details the maintenance, repair, and replacement of components, as well as the documentation required and the procedures to be followed to maintain the integrity of the TRUPACT-II container in accordance with OM-134, TRUPACT-II Container Operations and Maintenance Manual; and the TRUPACT-II Container Certificate of Compliance (Number 9218). The routine shipping and receiving inspections required by the Department of Transportation (DOT), Department of Energy (DOE), Nuclear Regulatory Commission (NRC) and other regulations are not addressed in this document. This document applies to all DOE shipping and receiving sites that use the TRUPACT-II containers

  3. Considerations of severe accidents in the design of Korean Next Generation Reactor

    International Nuclear Information System (INIS)

    Dong Wook Jerng; Choong Sup Byun

    1998-01-01

    The severe accident is one of the key issues in the design of Korean Next Generation Reactor (KNGR) which is an evolutionary type of pressurized water reactor. As IAEA recommends in TECDOC-801, the design objective of KNGR with regard to safety is provide a sound technical basis by which an imminent off-site emergency response to any circumstance could be practically unnecessary. To implement this design objective, probabilistic safety goals were established and design requirements were developed for systems to mitigate severe accidents. The basic approach of KNGR to address severe accidents is firstly prevent severe accidents by reinforcing its capability to cope with the design basis accidents (DBA) and further with some accidents beyond DBAs caused by multiple failures, and secondly mitigate severe accidents to ensure the retention of radioactive materials in the containment by providing mean to maintain the containment integrity. For severe accident mitigation, KNGR principally takes the concept of ex-vessel corium cooling. To implement this concept, KNGR is equipped with a large cavity and cavity flooding system connected to the in-containment refueling water storage tank. Other major systems incorporated in KNGR are hydrogen igniters and safety depressurization systems. In addition, the KNGR containment is designed to withstand the pressure and temperature conditions expected during the course of severe accidents. In this paper, the design features and status of system designs related with severe accidents will be presented. Also, R and D activities related to severe accident mitigation system design will be briefly described

  4. Spacetimes containing slowly evolving horizons

    International Nuclear Information System (INIS)

    Kavanagh, William; Booth, Ivan

    2006-01-01

    Slowly evolving horizons are trapping horizons that are ''almost'' isolated horizons. This paper reviews their definition and discusses several spacetimes containing such structures. These include certain Vaidya and Tolman-Bondi solutions as well as (perturbatively) tidally distorted black holes. Taking into account the mass scales and orders of magnitude that arise in these calculations, we conjecture that slowly evolving horizons are the norm rather than the exception in astrophysical processes that involve stellar-scale black holes

  5. Query containment in entity SQL

    OpenAIRE

    Rull Fort, Guillem; Bernstein, Philip A.; Garcia dos Santos, Ivo; Katsis, Yannis; Melnik, Sergey; Teniente López, Ernest

    2013-01-01

    We describe a software architecture we have developed for a constructive containment checker of Entity SQL queries defined over extended ER schemas expressed in Microsoft's Entity Data Model. Our application of interest is compilation of object-to-relational mappings for Microsoft's ADO.NET Entity Framework, which has been shipping since 2007. The supported language includes several features which have been individually addressed in the past but, to the best of our knowledge, they have not be...

  6. Aging characteristics of containment building and sensitivity on ultimate pressure capacity

    International Nuclear Information System (INIS)

    Seo, Jeong Moon; Choun, Young Sun; Choi, In Kil; Ha, Jae Joo

    1998-03-01

    For the reliable safety assessment of the containment building, structural and material conditions can be investigated in detail and pertinent assessment technologies have to be established. Also, an understanding on the aging-related degradations for the construction materials is required to predict long-term structural safety of the containment building. For the development of reliable aging prediction models, an extensive data base system related to aging properties of the containment building has to be prepared. The objectives of this research are to develop aging models representing long-term degradation of materials and a structural performance assessment program for containment building considering aging-related degradation. According to the results of sensitivity analysis, as the mechanical properties of the constituent materials degrade, the ultimate pressure capacity of containment building may decrease and severe damage may occur around the mid-level of the containment wall. (author). 28 refs., 11 tabs., 36 figs

  7. Weapon container catalog. Volumes 1 & 2

    Energy Technology Data Exchange (ETDEWEB)

    Brown, L.A.; Higuera, M.C.

    1998-02-01

    The Weapon Container Catalog describes H-gear (shipping and storage containers, bomb hand trucks and the ancillary equipment required for loading) used for weapon programs and for special use containers. When completed, the catalog will contain five volumes. Volume 1 for enduring stockpile programs (B53, B61, B83, W62, W76, W78, W80, W84, W87, and W88) and Volume 2, Special Use Containers, are being released. The catalog is intended as a source of information for weapon program engineers and also provides historical information. The catalog also will be published on the SNL Internal Web and will undergo periodic updates.

  8. Simplified safety and containment systems for the iris reactor

    International Nuclear Information System (INIS)

    Conway, L.E.; Lombardi, C.; Ricotti, M.; Oriani, L.

    2001-01-01

    The IRIS (International Reactor Innovative and Secure) is a 100 - 300 MW modular type pressurized water reactor supported by the U.S. DOE NERI Program. IRIS features a long-life core to provide proliferation resistance and to reduce the volume of spent fuel, as well as reduce maintenance requirements. IRIS utilizes an integral reactor vessel that contains all major primary system components. This integral reactor vessel makes it possible to reduce containment size; making the IRIS more cost competitive. IRIS is being designed to enhance reactor safety, and therefore a key aspect of the IRIS program is the development of the safety and containment systems. These systems are being designed to maximize containment integrity, prevent core uncover following postulated accidents, minimize the probability and consequences of severe accidents, and provide a significant simplification over current safety system designs. The design of the IRIS containment and safety systems has been identified and preliminary analyses have been completed. The IRIS safety concept employs some unique features that minimize the consequences of postulated design basis events. This paper will provide a description of the containment design and safety systems, and will summarize the analysis results. (author)

  9. T25 Qualification of old containers

    International Nuclear Information System (INIS)

    Hoerning, T.; Kasparek, E.M.; Voelzke, H.

    2013-01-01

    Radioactive waste for final disposal in the repository Konrad has to be packaged in licensed containers. The qualification of containers is performed according the respective standards and BfS requirements. The latter ones differ for different container types with respect to accident resistance based on the release behavior of the waste. Usually qualification of container types is performed before series production. The qualification of already produced containers (''old containers'') is significantly more complicated. The differences of quality assurance and materials properties of the older containers have to be evaluated. A main tool for this evaluation is the numerical re-calculation of accident scenarios for estimation of the safety margins. The validation of appropriate finite element models is available. Based on these date it might be possible to abandon additional drop tests with the existing old containers to prove the structural integrity.

  10. Sulfur-Containing Agrochemicals.

    Science.gov (United States)

    Devendar, Ponnam; Yang, Guang-Fu

    2017-10-09

    Modern agricultural chemistry has to support farmers by providing innovative agrochemicals. In this context, the introduction of sulfur atoms into an active ingredient is still an important tool in modulating the properties of new crop-protection compounds. More than 30% of today's agrochemicals contain at least one sulfur atom, mainly in fungicides, herbicides and insecticides. A number of recently developed sulfur-containing agrochemical candidates represent a novel class of chemical compounds with new modes of action, so we intend to highlight the emerging interest in commercially active sulfur-containing compounds. This chapter gives a comprehensive overview of selected leading sulfur-containing pesticidal chemical families namely: sulfonylureas, sulfonamides, sulfur-containing heterocyclics, thioureas, sulfides, sulfones, sulfoxides and sulfoximines. Also, the most suitable large-scale synthetic methods of the recently launched or provisionally approved sulfur-containing agrochemicals from respective chemical families have been highlighted.

  11. Reactor container structure

    International Nuclear Information System (INIS)

    Sato, Yoshimi; Fukuda, Yoshio.

    1993-01-01

    A main container of an FBR type reactor using liquid sodium as coolants is attached to a roof slug. The main container contains, as coolants, lower temperature sodium, and high temperature sodium above a reactor core and a partitioning plate. The main container has a structure comprising only longitudinal welded joints in parallel with axial direction in the vicinity of the liquid surface of high temperature sodium where a temperature gradient is steep and great thermal stresses are caused without disposing lateral welded joints in perpendicular to axial direction. Only the longitudinal welded joints having a great fatigue strength are thus disposed in the vicinity of the liquid surface of the high temperature sodium where axial thermal stresses are caused. This can improve reliability of strength at the welded portions of the main container against repeating thermal stresses caused in vicinity of the liquid surface of the main container from a view point of welding method. (I.N.)

  12. High security container

    International Nuclear Information System (INIS)

    Moreau, P.J.-M.; Monsterleet, G.A.

    1979-01-01

    This invention concerns containments, vessels or tanks for containing and protecting products or installations of various kinds, to be called by the general denomination 'containers'. Such products can be, inter alia, liquids such as natural gas, ammonia, vinyle chloride and hydrocarbons. Far from just forming simple means of storage, the containers used for this must now be capable of withstanding fire, sabotage for instance rocket fire, even impacts from aircraft, earthquakes and other aggressions of the same kind. The particular object of this invention is to create a container withstanding all these various agressions. It must also be considered that this container can not only be used for storing products or materials but also for enclosing particularly dangerous or delicate installations, such as nuclear or chemical reactors [fr

  13. Strategies for the prevention and mitigation of severe accidents

    International Nuclear Information System (INIS)

    Ader, C.; Heusener, G.; Snell, V.G.

    1999-01-01

    The currently operating nuclear power plants have, in general, achieved a high level of safety, as a result of design philosophies that have emphasized concepts such as defense-in-depth. This type of an approach has resulted in plants that have robust designs and strong containments. These designs were later found to have capabilities to protect the public from severe accidents (accidents more severe than traditional design basis in which substantial damage is done to the reactor core). In spite of this high level of safety, it has also been recognized that future plants need to be designed to achieve an enhanced level of safety, in particular with respect to severe accidents. This has led both regulatory authorities and utilities to develop guidance and/or requirements to guide plant designers in achieving improved severe accident performance through prevention and mitigation. The considerable research programs initiated after the TMI-2 accident have provided a large body of technical data, analytical methods, and the expertise necessary to provide for an understanding of a range of severe accident phenomena. This understanding of the ways severe accidents can progress and challenge containments, combined with the wide use of probabilistic safety assessments, have provided designers of evolutionary water cooled reactors opportunities to develop designs that minimize the challenges to the plant and to the public from severe accidents, including the development of accident management strategies intended to further reduce the risk of severe accidents. This paper describes some of the recent progress made in the understanding of severe accidents and related safety assessment methodology and how this knowledge has supported the incorporation of features into representative evolutionary designs that will prevent or mitigate many of the severe accident challenges present in current plants. (author)

  14. Severe Aplastic Anemia (SAA)

    Science.gov (United States)

    ... page Print this page My Cart Severe aplastic anemia (SAA) Severe aplastic anemia (SAA) is a disease ... leukemia (ALL) Other diseases What is severe aplastic anemia (SAA)? SAA is a bone marrow disease. The ...

  15. Pituitary tumors containing cholecystokinin

    DEFF Research Database (Denmark)

    Rehfeld, J F; Lindholm, J; Andersen, B N

    1987-01-01

    We found small amounts of cholecystokinin in the normal human adenohypophysis and therefore examined pituitary tumors from 87 patients with acromegaly, Cushing's disease, Nelson's syndrome, prolactinoma, or inactive pituitary adenomas. Five adenomas associated with Nelson's syndrome contained......'s disease and 7 acromegaly with adenomas containing ACTH. The cholecystokinin peptides from the tumors were smaller and less sulfated than cholecystokinin from normal pituitary glands. We conclude that ACTH-producing pituitary cells may also produce an altered form of cholecystokinin....

  16. Multicusp plasma containment apparatus

    International Nuclear Information System (INIS)

    Limpaecher, R.

    1980-01-01

    It has been discovered that plasma containment by a chamber having multi-pole magnetic cusp reflecting walls in combination with electronic injection for electrostatic containment provides the means for generating magnetic field free quiescent plasmas for practical application in ion-pumps, electronic switches, and the like. 1250 ''alnico v'' magnets 1/2 '' X 1/2 '' X 1 1/2 '' provide containment in one embodiment. Electromagnets embodying toroidal funneling extend the principle to fusion apparatus

  17. Radioactive material transporting container

    International Nuclear Information System (INIS)

    Watabe, Yukio.

    1990-01-01

    As a supporting member of a sealing container for containing spent fuels, etc., a straight pipe or a cylinder has been used. However, upon dropping test, the supporting member is buckled toward the central axis of a transporting container and a shock absorber is crushed in the axial direction to prevent its pushing force to the outer side, which may possibly hinder normal shock moderating function. Then, at least more than one-half of the supporting member is protruded radially to the outer side of the sealing container beyond the fixed portion with the sealed container, so that the member has a portion extended in the radial outside of the transporting container with an angle greater than the angle formed between a line connecting the outer circumference at the bottom of an outer cylinder with the gravitational center of the transporting container and the central axis of the transporting container. As a result, buckling of the supporting member toward the central axis of the transporting container upon dropping test can be prevented and the deformation of the shock absorber is neither not prevented to exhibit normal shock absorbing effect. This can improve the reliability and reduce the amount of shock absorbers. (N.H.)

  18. Inspection of Nuclear Power Plant Containment Structures

    Energy Technology Data Exchange (ETDEWEB)

    Graves, H.L.; Naus, D.J.; Norris, W.E.

    1998-12-01

    Safety-related nuclear power plant (NPP) structures are designed to withstand loadings from a number of low-probability external and interval events, such as earthquakes, tornadoes, and loss-of-coolant accidents. Loadings incurred during normal plant operation therefore generally are not significant enough to cause appreciable degradation. However, these structures are susceptible to aging by various processes depending on the operating environment and service conditions. The effects of these processes may accumulate within these structures over time to cause failure under design conditions, or lead to costly repair. In the late 1980s and early 1990s several occurrences of degradation of NPP structures were discovered at various facilities (e.g., corrosion of pressure boundary components, freeze- thaw damage of concrete, and larger than anticipated loss of prestressing force). Despite these degradation occurrences and a trend for an increasing rate of occurrence, in-service inspection of the safety-related structures continued to be performed in a somewhat cursory manner. Starting in 1991, the U.S. Nuclear Regulatory Commission (USNRC) published the first of several new requirements to help ensure that adequate in-service inspection of these structures is performed. Current regulatory in-service inspection requirements are reviewed and a summary of degradation experience presented. Nondestructive examination techniques commonly used to inspect the NPP steel and concrete structures to identify and quantify the amount of damage present are reviewed. Finally, areas where nondestructive evaluation techniques require development (i.e., inaccessible portions of the containment pressure boundary, and thick heavily reinforced concrete sections are discussed.

  19. Severe accident management guidelines tool

    International Nuclear Information System (INIS)

    Gutierrez Varela, Javier; Tanarro Onrubia, Augustin; Martinez Fanegas, Rafael

    2014-01-01

    Severe Accident is addressed by means of a great number of documents such as guidelines, calculation aids and diagnostic trees. The response methodology often requires the use of several documents at the same time while Technical Support Centre members need to assess the appropriate set of equipment within the adequate mitigation strategies. In order to facilitate the response, TECNATOM has developed SAMG TOOL, initially named GGAS TOOL, which is an easy to use computer program that clearly improves and accelerates the severe accident management. The software is designed with powerful features that allow the users to focus on the decision-making process. Consequently, SAMG TOOL significantly improves the severe accident training, ensuring a better response under a real situation. The software is already installed in several Spanish Nuclear Power Plants and trainees claim that the methodology can be followed easier with it, especially because guidelines, calculation aids, equipment information and strategies availability can be accessed immediately (authors)

  20. TRUPACT-II container maintenance program plan

    International Nuclear Information System (INIS)

    1990-11-01

    This document details the maintenance/repair and replacement of components, as well as the documentation required and the procedures to be followed to maintain the integrity of the TRUPACT-II container, in accordance with requirements of the TRUPACT-II Container Operations and Maintenance Manual, OM-134, the TRUPACT-II Container Safety Analysis Report (SARP), and the TRUPACT-II Container Certificate of Compliance (Number 9218). The routine shipping and receiving inspections required by the Department of Transportation (DOT), Department of Energy (DOE), Nuclear Regulatory Commission (NRC) and other regulations are not addressed in this document. This document applies to all DOE shipping and receiving sites that use the TRUPACT-II containers