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)

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

Containment Emergency Sump Performance. Technical findings related to Unresolved Safety Issue A-43. Revision 1

International Nuclear Information System (INIS)

1985-10-01

This report summarizes key technical findings related to Unresolved Safety Issue (USI) A-43, Containment Emergency Sump Performance. Both BWRs and PWRs are considered in this report. Emergency core cooling systems require a clean, reliable water source to maintain long-term recirculation following a loss-of-coolant accident (LOCA). PWRs rely on the containment emergency sump to provide such a water supply to residual heat removal pumps and containment spray pumps. BWRs rely on pump suction intakes in the suppression pool or wet well to provide water to residual heat removal and core spray systems. Thus, the technical findings in this report provide information on post-LOCA recirculation. These findings have been derived from extensive experimental studies, generic plant studies, and assessments of sumps used for long-term cooling. Results of hydraulic tests have shown that the potential for air ingestion is less severe than previously hypothesized. The effects of debris blockage on NPSH margin must be dealt with on a plant-specific basis. These findings have been used to develop revisions to Regulatory Guide 1.82 and Standard Review Plan Section 6.2.2 (NUREG-0800)

Identification of long-term containment/stabilization technology performance issues

International Nuclear Information System (INIS)

Matthern, G.E.; Nickelson, D.F.

1997-01-01

U.S. Department of Energy (DOE) faces a somewhat unique challenge when addressing in situ remedial alternatives that leave long-lived radionuclides and hazardous contaminants onsite. These contaminants will remain a potential hazard for thousands of years. However, the risks, costs, and uncertainties associated with removal and offsite disposal are leading many sites to select in situ disposal alternatives. Improvements in containment, stabilization, and monitoring technologies will enhance the viability of such alternatives for implementation. DOE's Office of Science and Technology sponsored a two day workshop designed to investigate issues associated with the long-term in situ stabilization and containment of buried, long-lived hazardous and radioactive contaminants. The workshop facilitated communication among end users representing most sites within the DOE, regulators, and technologists to define long-term performance issues for in situ stabilization and containment alternatives. Participants were divided into groups to identify issues and a strategy to address priority issues. This paper presents the results of the working groups and summarizes the conclusions. A common issue identified by the work groups is communication. Effective communication between technologists, risk assessors, end users, regulators, and other stakeholders would contribute greatly to resolution of both technical and programmatic issues

Heat transfer effects in vertically emplaced high level nuclear waste container

International Nuclear Information System (INIS)

Moujaes, S.F.; Lei, Y.M.

1994-01-01

Modeling free convection heat transfer in an cylindrical annular enclosure is still an active area of research and an important problem to be addressed in the high level nuclear waste repository. For the vertically emplaced waste container, the air gap which is between the container shell and the rock borehole, have an important role of dissipating heat to surrounding rack. These waste containers are vertically emplaced in the borehole 300 meters below ground, and in a horizontal grid of 30 x 8 meters apart. The borehole will be capped after the container emplacement. The expected initial heat generated is between 3--4.74 kW per container depending on the type of waste. The goal of this study is to use a computer simulation model to find the borehole wall, air-gap and the container outer wall temperature distributions

An experimental investigation on ground heat flow balance issue for a GCHP

OpenAIRE

Jiufa Chen; Hongqi Zheng; Qin Xue; Erming An; Weilai Qiao

2010-01-01

For a ground-coupled heat pump (GCHP), it is vital to keep the ground heat flow balanced in order to achieve sustainable energy-saving operation. However, the importance of this issue has not been well studied. Focused on the heat flow balance issue, this paper made an exclusive experimental study using a newly installed GCHP system with the designed cooling capacity 1960 kW and heating capacity 1590 kW. The GCHP system was equipped with a data acquisition system and had temperature sensors i...

Primary Issues of Mixed Convection Heat Transfer Phenomena

Energy Technology Data Exchange (ETDEWEB)

Chae, Myeong-Seon; Chung, Bum-Jin [Kyung Hee University, Yongin (Korea, Republic of)

2015-10-15

The computer code analyzing the system operating and transient behavior must distinguish flow conditions involved with convective heat transfer flow regimes. And the proper correlations must be supplied to those flow regimes. However the existing safety analysis codes are focused on the Light Water Reactor and they are skeptical to be applied to the GCRs (Gas Cooled Reactors). One of the technical issues raise by the development of the VHTR is the mixed convection, which occur when the driving forces of both forced and natural convection are of comparable magnitudes. It can be encountered as in channel of the stacked with fuel elements and a decay heat removal system and in VHTR. The mixed convection is not intermediate phenomena with natural convection and forced convection but independent complicated phenomena. Therefore, many researchers have been studied and some primary issues were propounded for phenomena mixed convection. This paper is to discuss some problems identified through reviewing the papers for mixed convection phenomena. And primary issues of mixed convection heat transfer were proposed respect to thermal hydraulic problems for VHTR. The VHTR thermal hydraulic study requires an indepth study of the mixed convection phenomena. In this study we reviewed the classical flow regime map of Metais and Eckert and derived further issues to be considered. The following issues were raised: (1) Buoyancy aided an opposed flows were not differentiated and plotted in a map. (2) Experimental results for UWT and UHF condition were also plotted in the same map without differentiation. (3) The buoyancy coefficient was not generalized for correlating with buoyancy coefficient. (4) The phenomenon analysis for laminarization and returbulization as buoyancy effects in turbulent mixed convection was not established. (5) The defining to transition in mixed convection regime was difficult.

Heat removal tests for pressurized water reactor containment spray by largescale facility

International Nuclear Information System (INIS)

Motoki, Y.; Hashimoto, K.; Kitani, S.; Naritomi, M.; Nishio, G.; Tanaka, M.

1983-01-01

Heat removal tests for pressurized water reactor (PWR) containment spray were carried out to investigate effectiveness of the depressurization by Japan Atomic Energy Research Institute model containment (7-m diameter, 20 m high, and 708-m 3 volume) with PWR spray nozzles. The depressurization rate is influenced by the spray heat transfer efficiency and the containment wall surface heat transfer coefficient. The overall spray heat transfer efficiency was investigated with respect to spray flow rate, weight ratio of steam/air, and spray height. The spray droplet heat transfer efficiency was investigated whether the overlapping of spray patterns gives effect or not. The effect was not detectable in the range of large value of steam/air, however, it was better in the range of small value of it. The experimental results were compared with the calculated results by computer code CONTEMPT-LT/022. The overall spray heat transfer efficiency was almost 100% in the containment pressure, ranging from 2.5 to 0.9 kg/cm 2 X G, so that the code was useful on the prediction of the thermal hydraulic behavior of containment atmosphere in a PWR accident condition

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)

Industrial implementation issues of Total Site Heat Integration

International Nuclear Information System (INIS)

Chew, Kew Hong; Klemeš, Jiří Jaromír; Wan Alwi, Sharifah Rafidah; Abdul Manan, Zainuddin

2013-01-01

Heat Integration has been a well-established energy conservation strategy in the industry. Total Site Heat Integration (TSHI) has received growing interest since its inception in the 90s. The methodology has been used with certain simplifications to solve TSHI problems. This paper investigates the main issues that can influence the practical implementation of TSHI in the industry. The main aim is to provide an assessment and possible guidance for future development and extension of the TSHI methodology from the industrial perspective. Several key issues have been identified as being of vital importance for the industries: design, operation, reliability/availability/maintenance, regulatory/policy and economics. Design issues to consider include plant layout, pressure drop, etc. For operation, issues such as startup and shutdown need to be considered. Reliability, availability and maintenance (RAM) are important as they directly affect the production. Relevant government policy and incentives are also important when considering the options for TSHI. Finally, a TSHI system needs to be economically viable. This paper highlights the key issues to be considered for a successful implementation of TSHI. The impacts of these issues on TS integration are summarised in a matrix, which forms a basis for an improved and closer-to-real-life implementation of the TSHI methodology. Highlights: ► Current TSHI methodology has been used for solving models with certain simplifications. ► Several issues that can influence practical implementation of TSHI are identified. ► Impacts of these issues on safety, environment and economics are evaluated. ► The findings form a basis for an improved and practical implementation of TSHI

Heat transfer coefficient for lead matrixing in disposal containers for used reactor fuel

International Nuclear Information System (INIS)

Mathew, P.M.; Taylor, M.; Krueger, P.A.

1985-02-01

In the Canadian Nuclear Fuel Waste Management Program, metal matrices with low melting points are being evaluated for their potential to provide support for the shell of disposal containers for used fuel, and to act as an additional barrier to the release of radionuclides. The metal matrix would be incorporated into the container by casting. To study the heat transfer processes during solidification, a steady-state technique was used, involving lead as the cast metal, to determine the overall heat transfer coefficient between the lead and some of the candidate container materials. The existence of an air gap between the cast lead and the container material appeared to control the overall heat transfer coefficient. The experimental observations indicated that the surface topography of the container material influences the heat transfer and that a smoother surface results in a greater heat transfer than a rough surface. The experimental results also showed an increasing heat transfer coefficient with increasing temperature difference across the container base plates; a model developed to base-plate bending can explain the observed results

Adiabatic equilibrium models for direct containment heating

International Nuclear Information System (INIS)

Pilch, M.; Allen, M.D.

1991-01-01

Probabilistic risk assessment (PRA) studies are being extended to include a wider spectrum of reactor plants than was considered in NUREG-1150. There is a need for simple direct containment heating (DCH) models that can be used for screening studies aimed at identifying potentially significant contributors to overall risk in individual nuclear power plants. This paper presents two adiabatic equilibrium models suitable for the task. The first, a single-cell model, places a true upper bound on DCH loads. This upper bound, however, often far exceeds reasonable expectations of containment loads based on CONTAIN calculations and experiment observations. In this paper, a two cell model is developed that captures the major mitigating feature of containment compartmentalization, thus providing more reasonable estimates of the containment load

Resolution of issues related to alternative RCS injection in the absence of containment sump recirculation

International Nuclear Information System (INIS)

Charles L Kling; Stephen S Barshay; Mathew C Jacob; Michael J Friedman

2005-01-01

Full text of publication follows: On June 9, 2003 the US NRC issued Bulletin No. 2003-01 that deals with the potential impact of debris blockage on containment sump recirculation at PWRs during a Loss-of-Coolant Accident (LOCA). In response to the bulletin, the Omaha Public Power District (OPPD) is in the process of developing procedural and operational strategies for their Fort Calhoun Station (FCS) to address the issues raised. Westinghouse provided engineering support to OPPD in identifying and resolving issues related to alternative means of supplying safety injection water to the reactor coolant system (RCS) in the absence of containment sump recirculation. Nuclear power plants are designed to protect the core following a LOCA by providing a continuous supply of cooling water to the core. In the long term, the Refueling Water Storage Tank (RWST) inventory will be depleted and core heat removal accomplished via recirculation of water previously injected into the Reactor Coolant System (RCS) and collected in the containment sump. Debris generated within the containment as a result of the impingement of fluid jets in the Zone of Influence (ZOI) of the RCS break and containment wash down may find its way into the containment sump. As the safety injection pumps take suction from the sump, in the recirculation mode of operation, the debris suspended in the sump water could begin to accumulate in the sump screen that is located in the recirculation path. Should sufficient debris accumulate on the sump screen, a flow blockage could potentially develop. This would result in insufficient safety injection pump NPSH, thereby impairing the recirculation mode of injection into RCS. Potential debris blockage and prevention of sump recirculation may be addressed by refilling the RWST with water and injecting this water directly into the core. This paper identifies and attempts to resolve several issues related to this alternative mode of RCS injection. In particular, the

Heat transfer in a fuel pin shipping container

International Nuclear Information System (INIS)

Ingham, J.G.

1980-01-01

Maximum cladding temperatures occur when the IDENT 1578 fuel pin shipping container is installed in the T-3 Cask. The maximum allowable cladding temperature of 800 0 F is reached when the rate of energy deposited in the 19-pin basket reaches 400 watts. Since 45% of the energy which is generated in the fuel escapes the 19-pin basket without being deposited, mostly gamma energy, the maximum allowable rate of heat generation is 400/.55 = 727 watts. Similarly, the maximum allowable cladding temperature of 800 0 F is reached when the rate of energy deposited in the 40-pin basket reaches 465 watts. Since 33% of the energy which is generated in the fuel escapes the 40-pin basket without being deposited, mostly gamma energy, the maximum allowable rate of heat generation is 465/.66 = 704 watts. The IDENT 1578 fuel pin shipping container therefore meets its thermal design criteria. IDENT 1578 can handle fuel pins with a decay heat load of 600 watts while maintaining the maximum fuel pin cladding temperature below 800 0 F. The emissivities which were determined from the test results for the basket tubes and container are relatively low and correspond to new, shiny conditions. As the IDENT 1578 container is exposed to high temperatures for extended periods of time during the transportation of fuel pins, the emissivities will probably increase. This will result in reduced temperatures

Heat removing device for reactor container

International Nuclear Information System (INIS)

Hisamochi, Kohei; Matsumoto, Tomoyuki; Matsumoto, Masayoshi; Sato, Ken-ichi.

1996-01-01

A recycling loop for reactor water is disposed in a reactor pressure vessel of a BWR type reactor. Extracted reactor water from the recycling loop passes through a extracted reactor water pipeline and flows into a reactor coolant cleanup system. A pipeline for connecting the extracted reactor water pipeline and a suppression pool is disposed, and a discharged water pressurizing pump is disposed to the pipeline. Upon occurrence of emergency, discharged water from the suppression pool is pressurized by a discharged water pressurizing pump and sent to a reactor coolant cleanup system. The discharged water is cooled while passing through a sucking water cooling portion of a regenerative heat exchanger and a non-regenerative heat exchanger. Then, it is sent to a feed water pipeline passing a bypass line of a filtering desalter and a bypass line of the sucked water cooling portion of the regenerative heat exchanger, injected to the inside of the pressure vessel to cool the reactor core and remove after-heat. Then, it cools the inside of the reactor container together with coolants flown out of the pressure vessel and then returns to the suppression pool. (I.N.)

9 CFR 355.25 - Canning with heat processing and hermetically sealed containers; closures; code marking; heat...

Science.gov (United States)

2010-01-01

... 9 Animals and Animal Products 2 2010-01-01 2010-01-01 false Canning with heat processing and hermetically sealed containers; closures; code marking; heat processing; incubation. 355.25 Section 355.25... IDENTIFICATION AS TO CLASS, QUALITY, QUANTITY, AND CONDITION Inspection Procedure § 355.25 Canning with heat...

Heat-processing method and facility for helium-containing metal material

International Nuclear Information System (INIS)

Kato, Takahiko; Kodama, Hideyo; Matsumoto, Toshimi; Aono, Yasuhisa; Nagata, Tetsuya; Hattori, Shigeo; Kaneda, Jun-ya; Ono, Shigeki.

1996-01-01

Electric current is supplied to an objective portion of a He-containing metal material to be applied with heat processing without causing melting, to decrease the He content of the portion. Subsequently, the defect portion of the tissues of the He-containing metal is modified by heating the portion with melting. Since electric current can be supplied to the metal material in a state where the metal material is heated and the temperature thereof is elevated, an effect of further reducing the He content can be obtained. Further, if the current supply and/or the heating relative to the metal material is performed in a vacuum or inert gas atmosphere, an effect of reducing the degradation of the surface of the objective portion to be supplied with electric current can be obtained. (T.M.)

The technological raw material heating furnaces operation efficiency improving issue

Science.gov (United States)

Paramonov, A. M.

2017-08-01

The issue of fuel oil applying efficiency improving in the technological raw material heating furnaces by means of its combustion intensification is considered in the paper. The technical and economic optimization problem of the fuel oil heating before combustion is solved. The fuel oil heating optimal temperature defining method and algorithm analytically considering the correlation of thermal, operating parameters and discounted costs for the heating furnace were developed. The obtained optimization functionality provides the heating furnace appropriate thermal indices achievement at minimum discounted costs. The carried out research results prove the expediency of the proposed solutions using.

Twin cities institutional issues study cogenerated hot water district heating

Energy Technology Data Exchange (ETDEWEB)

Sundberg, R. E.; Leas, R.; Kolb, J. O.

1979-01-01

Community district heating, utilizing hot water produced through electrical/thermal cogeneration, is seen as an integral part of Minnesota's Energy Policy and Conservation Plan. Several studies have been conducted which consider the technical and institutional issues affecting implementation of cogenerated district heating in the Minneapolis and St. Paul Metropolitan Area. The state of the technical art of cogenerated hot water district heating is assumed to be transferable from European experience. Institutional questions relating to such factors as the form of ownership, financing, operation, regulation, and product marketability cannot be transferred from the European experience, and have been the subject of an extensive investigation. The form and function of the Institutional Issues Study, and some of the preliminary conclusions and recommendations resulting from the study are discussed.

Heat transfer effects in vertically emplaced high level nuclear waste container

International Nuclear Information System (INIS)

Moujaes, S.F.; Lei, Y.M.

1994-01-01

Modeling free convection heat transfer in a cylindrical annular enclosure is still an active area of research and an important problem to be addressed in the high level nuclear waste repository. For the vertically emplaced waste container, the air gap which is between the container shell and the rock borehole, have an important role of dissipating heat to surrounding rock. These waste containers are vertically emplaced in the borehole 300 meters just below ground, and in a horizontal grid of 30 x 8 meters apart. The borehole will be capped after the container emplacement. The expected initial heat generated is between 3-4.74 kW per container depending on the type of waste. The goal of this study is to use a computer simulation model to find the borehole wall, air-gap and the container outer wall temperature distributions. The borehole wall temperature history has been found in the previous study, and was estimated to reach a maximum temperature of about 218 degrees C after 18 years from the emplacement. The temperature history of the rock surface is then used for the air-gap simulation. The problem includes convection and radiation heat transfer in a vertical enclosure. This paper will present the results of the convection in the air-gap over one thousand years after the containers' emplacement. During this long simulation period it was also observed that a multi-cellular air flow pattern can be generated in the air gap

Comparison of CONTAIN and TCE calculations for direct containment heating of Surry

International Nuclear Information System (INIS)

Washington, K.E.; Stuart, D.S.

1996-01-01

This paper presents the results of several CONTAIN code calculations used to model direct containment heating (DCH) loads for the Surry plant. The results of these calculations are compared with the results obtained using the two-cell equilibrium (TCE) model for the same set of initial and boundary conditions. This comparison is important because both models have been favorably validated against the available DCH database, yet there are potentially important modeling differences. The comparisons are to quantitatively assess the impact of these differences. A major conclusion of this study is that, for the accident conditions studied and for a broad range of sensitivity cases, the peak pressures predicted by both TCE and CONTAIN are well below the failure pressure for the Surry containment. (orig.)

Heat removing device for nuclear reactor container facility

Energy Technology Data Exchange (ETDEWEB)

Tateno, Seiya; Tominaga, Kenji; Iwata, Yasutaka; Kinoshita, Shoichiro; Niino, Tsuyoshi

1994-09-30

A pressure suppression chamber incorporating pool water is disposed inside of a reactor container for condensating steams released to a dry well upon occurrence of abnormality. A pool is disposed at the outer circumference of the pressure suppression chamber having a steel wall surface of the reactor container as a partition wall. The outer circumferential pool is in communication with ocean by way of a lower communication pipeline and an upper communication pipeline. During normal plant operation state, partitioning valves disposed respectively to the upper and lower communication pipelines are closed, so that the outer circumferential pool is kept empty. After occurrence loss of coolant accident, steams generated by after-heat of the reactor core are condensated by pool water of the pressure suppression chamber, and the temperature of water in the pressure suppression chamber is gradually elevated. During the process, the partition valves of the upper and lower communication pipelines are opened to introduce cold seawater to the outer circumferential pool. With such procedures, heat of the outer circumferential pool is released to the sea by natural convection of seawater, thereby enabling to remove residual heat without dynamic equipments. (I.N.).

Central unresolved issues in thermal energy storage for building heating and cooling

Energy Technology Data Exchange (ETDEWEB)

Swet, C.J.; Baylin, F.

1980-07-01

This document explores the frontier of the rapidly expanding field of thermal energy storage, investigates unresolved issues, outlines research aimed at finding solutions, and suggests avenues meriting future research. Issues related to applications include value-based ranking of storage concepts, temperature constraints, consistency of assumptions, nomenclature and taxonomy, and screening criteria for materials. Issues related to technologies include assessing seasonal storage concepts, diurnal coolness storage, selection of hot-side storage concepts for cooling-only systems, phase-change storage in building materials, freeze protection for solar water heating systems, and justification of phase-change storage for active solar space heating.

LeBreton Flats district heating: Institutional issues

Energy Technology Data Exchange (ETDEWEB)

Gerow, L.B.; Bernstein, H.

1981-12-01

Despite the growing popularity in Europe of district heating as an alternative to conventional systems, a number of institutional barriers blocked the road to its successful implementation in Canada. These hurdles include questions of ownership, operation/administration, consumer accountability, financial arrangements, and general consumer acceptance of a new technology. In deciding on ownership, the major issues were legal. Provincial laws to be complied with included The Municipal Franchises Act, The Public Utilities Act, The Municipal Act and The Ontario Municipal Board Act. The operation and administration of the project were also beset with legal complications. The relevant laws were The Boilers and Pressure Vessels Act, and The Environmental Assessment Act. How to charge individual users (consumer accountability) raised a number of technical problems including metering and rate-setting for hot water distribution, and heat borrowing between units. The report recommends that groups planning to implement district heating are advised to allow plenty of lead time for obtaining approvals and franchising arrangements so as not to jeopardize construction schedules. 3 refs.

Heating and cooling device for use in the vacuum container of a thermonuclear device

International Nuclear Information System (INIS)

Morita, Hiroaki; Onozuka, Masanori; Fukui, Hiroshi.

1986-01-01

Purpose: To prevent the generation of great temperature difference within a hollow doughnuts-shaped space of the torus vacuum container of a tokamak type thermonuclear reactor, as well as effectively eliminate the local injection of heat to the vacuum container. Constitution: A hollow doughnuts-like space is formed between the inner wall and the double outer wall of a vacuum container main body, which is divided into a plurality of regions by partition plates extended in the toroidal direction. An input/output header is disposed in adjacent with each of the partition plates for inputting/outputting heat medium. Further, heat medium inlet/outlets are disposed to define two flow channels on every one-half circumference. This enables to reduce the temperature difference of the heat medium between the inlet and the outlet by the shortening of the flow channel length and heating or cooling can be performed without causing unevenness in the temperature distribution of the vacuum container. (Horiuchi, T.)

Containment response and radiological release for a TMLB' accident sequence in a large dry containment

International Nuclear Information System (INIS)

Gasser, R.D.; Bieniarz, P.P.; Tills, J.L.

1987-01-01

An analysis has been performed for the Bellefonte Pressurized Water Reactor (PWR) Unit 1 to determine the containment loading and the radiological releases into the environment from a station blackout accident. A number of issues have been addressed in this analysis, which include the effects of direct heating on containment loading and the effects of fission product heating and natural convection on releases from the primary system. The results indicate that direct heating, which involves more than about 50% of the core, may fail the Bellefonte containment, but natural convection in the Reactor Coolant System (RCS) may lead to overheating and failure of the primary system piping before core slump, thus, eliminating or mitigating direct heating. Releases from the primary system are significantly increased before vessel breach, due to natural circulation, and after vessel breach, due to reevolution of retained fission products by fission product heating of RCS structures. (orig.)

Thermal Performance and Operation Limit of Heat Pipe Containing Neutron Absorber

Energy Technology Data Exchange (ETDEWEB)

Kim, Kyung Mo; Jeong, Yeong Shin; Kim, In Guk; Bang, In Choel [UNIST, Ulsan (Korea, Republic of)

2015-05-15

Recently, passive safety systems are under development to ensure the core cooling in accidents involving impossible depressurization such as station blackout (SBO). Hydraulic control rod drive mechanisms, passive auxiliary feedwater system (PAFS), Passive autocatalystic recombiner (PAR), and so on are types of passive safety systems to enhance the safety of nuclear power plants. Heat pipe is used in various engineering fields due to its advantages in terms of easy fabrication, high heat transfer rate, and passive heat transfer. Also, the various concepts associated with safety system and heat transfer using the heat pipe were developed in nuclear engineering field.. Thus, our group suggested the hybrid control rod which combines the functions of existing control rod and heat pipe. If there is significant temperature difference between active core and condenser, the hybrid control rod can shutdown the nuclear fission reaction and remove the decay heat from the core to ultimate heat sink. The unique characteristic of the hybrid control rod is the presence of neutron absorber inside the heat pipe. Many previous researchers studied the effect of parameters on the thermal performance of heat pipe. However, the effect of neutron absorber on the thermal performance of heat pipe has not been investigated. Thus, the annular heat pipe which contains B{sub 4}C pellet in the normal heat pipe was prepared and the thermal performance of the annular heat pipe was studied in this study. Hybrid control rod concept was developed as a passive safety system of nuclear power plant to ensure the safety of the reactor at accident condition. The hybrid control rod must contain the neutron absorber for the function as a control rod. So, the effect of neutron absorber on the thermal performance of heat pipe was experimentally investigated in this study. Temperature distributions at evaporator section of annular heat pipe were lower than normal heat pipe due to the larger volume occupied by

Transition to chaos in a square enclosure containing internal heat sources

Energy Technology Data Exchange (ETDEWEB)

Baytas, A.C. [Institute For Nuclear Energy, Istanbul (Turkey)

1995-09-01

A numerical investigation is performed to study the transition from steady to chaotic flow of a fluid confined in a two-dimensional square cavity. The cavity has rigid walls of constant temperature containing uniformly distributed internal heat source. Effects of the Rayleigh number of flow and heat transfer rates are studied. In addition to, same problem is solved for sinusoidally changing internal heat source to show its effect on the flow model and heat transfer of the enclosures. Details of oscillatory solutions and flow bifurcations are presented.

Effect of heat treatment duration on phase separation of sodium borosilicate glass, containing copper

International Nuclear Information System (INIS)

Shejnina, T.G.; Gutner, S.Kh.; Anan'in, N.I.

1989-01-01

The effect of heat treatment duration on phase separation of sodium borosilicate (SBS) glass, containing copper is studied. It is stated that phase separation close to equilibrium one is attained under 12 hours of heat treatment of SBS glass containing copper

DHCVIM - a direct heating containment vessel interactions module: applications to Sandia National Laboratories Surtsey experiments

International Nuclear Information System (INIS)

Ginsberg, T.; Tutu, N.K.

1987-01-01

Direct containment heating is the mechanism of severe nuclear reactor accident containment loading that results from transfer of thermal and chemical energy from high-temperature, finely divided, molten core material to the containment atmosphere. The direct heating containment vessel interactions module (DHCVIM) has been developed at Brookhaven National Laboratory to model the mechanisms of containment loading resulting from the direct heating accident sequence. The calculational procedure is being used at present to model the Sandia National Laboratories one-tenth-scale Surtsey direct containment heating experiments. The objective of the code is to provide a test bed for detailed modeling of various aspects of the thermal, chemical, and hydrodynamic interactions that are expected to occur in three regions of a containment building: reactor cavity, intermediate subcompartments, and containment dome. Major emphasis is placed on the description of reactor cavity dynamics. This paper summarizes the modeling principles that are incorporated in DHCVIM and presents a prediction of the Surtsey Test DCH-2 that was made prior to execution of the experiment

Condensation heat transfer with noncondensable gas for passive containment cooling of nuclear reactors

Energy Technology Data Exchange (ETDEWEB)

Leonardi, Tauna [Schlumberger, 14910 Airline Rd., Rosharon, TX 77583 (United States)]. E-mail: Tleonardi@slb.com; Ishii, Mamoru [School of Nuclear Engineering, Purdue University, West Lafayette, IN 47907 (United States)]. E-mail: Ishii@ecn.purdue.edu

2006-09-15

Noncondensable gases that come from the containment and the interaction of cladding and steam during a severe accident deteriorate a passive containment cooling system's performance by degrading the heat transfer capabilities of the condensers in passive containment cooling systems. This work contributes to the area of modeling condensation heat transfer with noncondensable gases in integral facilities. Previously existing correlations and models are for the through-flow of the mixture of steam and the noncondensable gases and this may not be applicable to passive containment cooling systems where there is no clear passage for the steam to escape. This work presents a condensation heat transfer model for the downward cocurrent flow of a steam/air mixture through a condenser tube, taking into account the atypical characteristics of the passive containment cooling system. An empirical model is developed that depends on the inlet conditions, including the mixture Reynolds number and noncondensable gas concentration.

Research on heat and mass transfer model for passive containment cooling system

International Nuclear Information System (INIS)

Jiang Xiaowei; Yu Hongxing; Sun Yufa; Huang Daishun

2013-01-01

Different with the traditional dry style containment design without external cooling, the PCCS design increased the temperature difference between the wall and the containment atmosphere significantly, and also the absolute temperature of the containment surfaces will be lower, affecting properties relevant in the condensation process. A research on the heat and mass transfer model has been done in this paper, especially the improvement on the condensation and evaporation model in the presence of noncondensable gases. Firstly, the Peterson's diffusion layer model was proved to equivalent to the stagnant film model adopted by CONTAIN code using the Clausius-Clapeyron equation, then a factor which can be used to stagnant film model was derived from the comparison between the Y.Liao's generalized diffusion layer model and the Peterson's diffusion layer model. Finally, the model in CONTAIN code used to compute the condensation and evaporation mass flux was modified using the factor, and the Wisconsin condensation tests and Westinghouse film evaporation on heated plate tests were simulated which had proved the improved model can predict more closer value of the heat and mass transfer coefficient to experimental value than original model. (authors)

Dynamic response of the target container under pulsed heating

Energy Technology Data Exchange (ETDEWEB)

Liping Ni [Paul Scherrer Inst. (PSI), Villigen (Switzerland)

1997-09-01

The structural mechanics of a liquid target container for pulsed spallation sources have been simulated using both a commercial code and a PSI-developed program. Results from the transient thermal-structural analysis showed that, due to inertia effects, the dynamic stress in the target container is contributed mainly from direct heating in the initial time stage, and later from the pressure wave in the target liquid once it reaches the wall. (author) figs., tab., refs.

Experimental simulation of corium dispersion phenomena in direct containment heating

International Nuclear Information System (INIS)

Wu, Q.

1996-01-01

In a direct containment heating (DCH) accident scenario, the degree of corium dispersion is one of the most significant factors responsible for the reactor containment heating and pressurization. To study the mechanisms of the corium dispersion phenomenon, a DCH separate effect test facility of 1:10 linear scale for Zion PWR geometry is constructed. Experiments are carried out with air-water and air-woods metal simulating steam and molten core materials. The physical process of corium dispersion is studied in detail through various instruments, as well as with flow visualization at several locations. The accident transient begins with the liquid jet discharge at the bottom of the reactor pressure vessel. Once the jet impinges on the cavity bottom floor, it immediately spreads out and moves rapidly to the cavity exit as a film flow. Part of the discharged liquid flows out of the cavity before gas blowdown, and the rest is subjected to the entrainment process due to the high speed gas stream. The liquid film and droplet flows from the reactor cavity will then experience subcompartment trapping and re-entrainment. Consequently, the dispersed liquid droplets that follow the gas stream are transported into the containment atmosphere, resulting in containment heating and pressurization in the prototypic condition. Comprehensive measurements are obtained in this study, including the liquid jet velocity, liquid film thickness and velocity transients in the test cavity, gas velocity and velocity profile in the cavity, droplet size distribution and entrainment rate, and the fraction of dispersed liquid in the containment building. These data are of great importance for better understanding of the corium dispersion mechanisms. (orig.)

The Heat Is On: Decision-Maker Perspectives on When and How to Issue a Heat Warning

Science.gov (United States)

O'Neill, M.; Sampson, N.; McCormick, S.; Rood, R. B.; Buxton, M.; Ebi, K. L.; Gronlund, C. J.; Zhang, K.; Catalano, L.; White-Newsome, J. L.; Conlon, K. C.; Parker, E. A.

2011-12-01

To better understand how to prevent illness and deaths during hot weather, particularly among at-risk populations, we conducted a study in Detroit, Michigan; Phoenix, Arizona; New York, New York, and Philadelphia, Pennsylvania. Our aims were to characterize and better understand how heatwave and health early warning systems (HHWS) and related prevention and sustainability programs can be more widely and effectively implemented. Specifically, we here report on the scientific evidence, expert judgments and the process used in deciding to trigger a HHWS and activate public health and social services interventions. We conducted interviews with public officials who decide if and when heat advisories/warnings are issued. After transcribing the interviews, we used a qualitative analysis software, QSR NVivo 9.0, to assign codes to portions of text from each transcript and allow analysis of information with common themes across the data. For example, several sentences in a transcript discussing a heat index might be coded as 'definition of heat wave'. A common theme across cities was that deciding what type of weather is dangerous to health is not straightforward. The time in season that heat occurs; the duration of the heat; the level of humidity and other meteorological factors; the extent to which temperatures drop at night, allowing people to cool off; and prevailing weather conditions all play a role. A single 'safe' threshold is unrealistic because people's individual sensitivity, housing, surrounding environments, behaviors, and access to air conditioning can differ greatly. However, choices must be made as to the trigger for the HHWS. Although quantitative analysis with health data (mortality, hospital admissions) can inform the design of the triggers, historical analysis has limitations, and decisions to issue heat warnings are sometimes related to planned activities, such as parades or fairs, that may expose large numbers of people to heat. The HHWS approach

The inaccuracy of heat transfer characteristics for non-insulated and insulated spherical containers neglecting the influence of heat radiation

International Nuclear Information System (INIS)

Wong, King-Leung; Salazar, Jose Luis Leon; Prasad, Leo; Chen, Wen-Lih

2011-01-01

In this investigation, the differences of heat transfer characteristics for insulated and non-insulated spherical containers between considering and neglecting the influence of heat radiation are studied by the simulations in some practical situations. It is found that the heat radiation effect cannot be ignored in conditions of low ambient convection heat coefficients (such ambient air) and high surface emissivities, especially for the non-insulated and thin insulated cases. In most practical situations when ambient temperature is different from surroundings temperature and the emissivity of insulation surface is different from that of metal wall surface, neglecting heat radiation will result in inaccurate insulation effect and heat transfer errors even with very thick insulation. However, the insulation effect considering heat radiation will only increase a very small amount after some dimensionless insulated thickness (such insulation thickness/radius ≥0.2 in this study), thus such dimensionless insulated thickness can be used as the optimum thickness in practical applications. Meanwhile, wrapping a material with low surface emissivity (such as aluminum foil) around the oxidized metal wall or insulation layer (always with high surface emissivity) can achieve very good insulated effect for the non-insulated or thin insulated containers.

Mobile heat storage containers and their transport by rail or road

Energy Technology Data Exchange (ETDEWEB)

Goldenberg, Philipp

2013-10-15

Mobile heat storage containers are capable of making a contribution to the meaningful use of energy which is needed for use at a location other than where it originates. The study presented in this report outlines the technology of mobile heat storage and analyses an example of its transport by rail or road. (orig.)

The prediction of direct containment heating

International Nuclear Information System (INIS)

Yan, H.; California Univ., Santa Barbara, CA; Theofanous, T.G.; California Univ., Santa Barbara, CA

1996-01-01

A simple analytical model is proposed and shown to capture the essence of the direct containment heating phenomenon. The model is based on assuming thermal/chemical equilibrium in the melt dispersal (flow) process, and separation of the melt out of this 'equilibrium steam' in the intermediate compartment. The model reveals a natural scale (hence named the 'DCH scale') for the DCH phenomenon, and the results are in very good agreement with the integral effects tests series. On this basis, reactor predictions can be made quite simply, provided that the DCH scale for the particular condition of interest is known. This prediction of DCH scale is also addressed by a scaling approach that is shown to be consistent with the experimental data. Finally, reactor predictions (of DCH loads) are also included in generalized terms convenient for use under a wide variety of conditions. In general, the results appear to be well within the structural capability of large dry containments. (orig.)

Heat transfer fluids containing nanoparticles

Science.gov (United States)

Singh, Dileep; Routbort, Jules; Routbort, A.J.; Yu, Wenhua; Timofeeva, Elena; Smith, David S.; France, David M.

2016-05-17

A nanofluid of a base heat transfer fluid and a plurality of ceramic nanoparticles suspended throughout the base heat transfer fluid applicable to commercial and industrial heat transfer applications. The nanofluid is stable, non-reactive and exhibits enhanced heat transfer properties relative to the base heat transfer fluid, with only minimal increases in pumping power required relative to the base heat transfer fluid. In a particular embodiment, the plurality of ceramic nanoparticles comprise silicon carbide and the base heat transfer fluid comprises water and water and ethylene glycol mixtures.

Using containment analysis to improve component cooling water heat exchanger limits

International Nuclear Information System (INIS)

Da Silva, H.C.; Tajbakhsh, A.

1995-01-01

The Comanche Peak Steam Electric Station design requires that exit temperatures from the Component Cooling Water Heat Exchanger remain below 330.37 K during the Emergency Core Cooling System recirculation stage, following a hypothetical Loss of Coolant Accident (LOCA). Due to measurements indicating a higher than expected combination of: (a) high fouling factor in the Component Cooling Water Heat Exchanger with (b) high ultimate heat sink temperatures, that might lead to temperatures in excess of the 330.37 K limit, if a LOCA were to occur, TUElectric adjusted key flow rates in the Component Cooling Water network. This solution could only be implemented with improvements to the containment analysis methodology of record. The new method builds upon the CONTEMPT-LT/028 code by: (a) coupling the long term post-LOCA thermohydraulics with a more detailed analytical model for the complex Component Cooling Water Heat Exchanger network and (b) changing the way mass and energy releases are calculated after core reflood and steam generator energy is dumped to the containment. In addition, a simple code to calculate normal cooldowns was developed to confirm RHR design bases were met with the improved limits

Hydrogen combustion issues and containment integrity

International Nuclear Information System (INIS)

Tennankore, K.N.; Koroll, G.W.; Kumar, R.K.; Lam, A.H.T.; Chan, C.K.; Wren, D.J.

1987-01-01

This report reviews recent results from the research program at the Whiteshell Nuclear Research Establishment aimed at resolving hydrogen combustion issues relevant to reactor containment integrity. The areas considered are hydrogen mixing behaviour, flame propagation, pressure transients caused by deflagration, flame acceleration and transition to detonation. Small-scale tests carried out with helium/CO 2 systems indicate that the presence of a recirculation path will enhance buoyancy-driven mixing of hydrogen-steam jets with air in the containment. An extensive database has been assembled for the laminar burning velocity of hydrogen-air-steam mixtures. Work is now beginning on obtaining burning velocities under turbulent conditions likely to prevail in containments. Turbulent and vented combustion data obtained in a 2.3-m-diameter spherical vessel have been analyzed. The results indicate that the pressure predictions of a one-dimensional model can be matched with the data if different effective burning velocities are used. Obstacle-induced flame acceleration and transition to detonation have been studied in a small-scale test facility to determine conditions required for transition to detonation in containments. For a repeated obstacle configuration, the obstacle blockage ratio required to cause transition (deflagation to detonation or detonation to deflagration) can be correlated well with detonation cell size. Cell sizes for mixtures of interest are being determined in our laboratory to complement available data. The transition-to-detonation process has been visualized in the above tests by Schlieren photography. The eventual objective of this work is to identify the local conditions controlling the transition process

Development of Latent Heat Storage Phase Change Material Containing Plaster

Directory of Open Access Journals (Sweden)

Diana BAJARE

2016-05-01

Full Text Available This paper reviews the development of latent heat storage Phase Change Material (PCM containing plaster as in passive application. Due to the phase change, these materials can store higher amounts of thermal energy than traditional building materials and can be used to add thermal inertia to lightweight constructions. It was shown that the use of PCMs have advantages stabilizing the room temperature variations during summer days, provided sufficient night ventilation is allowed. Another advantage of PCM usage is stabilized indoor temperature on the heating season. The goal of this study is to develop cement and lime based plaster containing microencapsulated PCM. The plaster is expected to be used for passive indoor applications and enhance the thermal properties of building envelope. The plaster was investigated under Scanning Electron Microscope and the mechanical, physical and thermal properties of created plaster samples were determined.

Experimental study on the heat transfer characteristics of a nuclear reactor containment wall cooled by gravitationally falling water

Science.gov (United States)

Pasek, Ari D.; Umar, Efrison; Suwono, Aryadi; Manalu, Reinhard E. E.

2012-06-01

Gravitationally falling water cooling is one of mechanism utilized by a modern nuclear Pressurized Water Reactor (PWR) for its Passive Containment Cooling System (PCCS). Since the cooling is closely related to the safety, water film cooling characteristics of the PCCS should be studied. This paper deals with the experimental study of laminar water film cooling on the containment model wall. The influences of water mass flow rate and wall heat rate on the heat transfer characteristic were studied. This research was started with design and assembly of a containment model equipped with the water cooling system, and calibration of all measurement devices. The containment model is a scaled down model of AP 1000 reactor. Below the containment steam is generated using electrical heaters. The steam heated the containment wall, and then the temperatures of the wall in several positions were measure transiently using thermocouples and data acquisition. The containment was then cooled by falling water sprayed from the top of the containment. The experiments were done for various wall heat rate and cooling water flow rate. The objective of the research is to find the temperature profile along the wall before and after the water cooling applied, prediction of the water film characteristic such as means velocity, thickness and their influence to the heat transfer coefficient. The result of the experiments shows that the wall temperatures significantly drop after being sprayed with water. The thickness of water film increases with increasing water flow rate and remained constant with increasing wall heat rate. The heat transfer coefficient decreases as film mass flow rate increase due to the increases of the film thickness which causes the increasing of the thermal resistance. The heat transfer coefficient increases slightly as the wall heat rate increases. The experimental results were then compared with previous theoretical studied.

Studies of heat transfer having relevance to nuclear reactor containment cooling by buoyancy-driven air flow

Energy Technology Data Exchange (ETDEWEB)

Jackson, J. D.; Li, J.; Wang, J. [The Univ., of Manchester, Manchester (United Kingdom)

2003-07-01

Two separate effects experiments concerned with buoyancy-influenced convective heat transfer in vertical passages which have relevance to the problem of nuclear reactor containment cooling by means of buoyancy-driven airflow are described. A feature of each is that local values of heat transfer coefficient are determined on surfaces maintained at uniform temperature. Experimental results are presented which highlight the need for buoyancy-induced impairment of turbulent convective heat transfer to be accounted for in the design of such passive cooling systems. A strategy is presented for predicting the heat removal by combined convective and radiative heat transfer from a full scale nuclear reactor containment shell using such experimental results.

Effect of heating oils and fats in containers of different materials on their trans fatty acid content.

Science.gov (United States)

Kala, A L Amrutha; Joshi, Vishal; Gurudutt, K N

2012-08-30

The nature of the container material and temperature employed for deep-frying can have an influence on the development of trans fatty acids (TFAs) in the fat used. The present study was undertaken to determine the effect of heating vegetable oils and partially hydrogenated vegetable fats with different initial TFA content in stainless steel, Hindalium (an aluminium alloy), cast iron and glass containers. Ground nut oil (oil 1), refined, bleached and deodorised (RBD) palmolein (oil 2) and two partially hydrogenated vegetable oils with low (fat 1) and high (fat 2) TFA content were uniformly heated at 175-185 °C over a period of 12 h. An increase in TFA content to 20 g kg⁻¹ was observed in oil 2 in the cast iron container, while a decrease in TFA content of 20-30 g kg⁻¹ was observed in fat 2 in all containers. The heating process of fats and oils also led to an increase in Butyro refractometer reading and colour values. This study showed that the TFA 18:1t content of oil 1, oil 2 and fat 1 increased with repeated or prolonged heating. The cast iron container showed the highest increase in TFA 18:1t for RBD palmolein (oil 2). The amount of linoleic acid trans isomers formed in the heating process was negligible. Fat 2 with high initial TFA content showed a decrease in TFA 18:1 and 18:2 on heating in all containers. Oils heated in glass and stainless steel containers showed less TFA 18:1t formation. Copyright © 2012 Society of Chemical Industry.

Initial performance assessment of the Westinghouse AP600 containment design and related safety issues

International Nuclear Information System (INIS)

Nicolette, V.F.; Washington, K.E.; Tills, J.L.

1991-01-01

This work summarizes the Westinghouse AP600 advanced reactor design assessment calculations performed to date with the CONTAIN code. Correlations for modeling the important heat transfer phenomena are discussed as well. A CONTAIN model of the AP600 was constructed for design basis accident (DBA) calculations. Insights gained from modeling of the smaller-scale Westinghouse Integral Test Facility were incorporated in the development of the AP600 model. The results of the DBA calculations are compared to the results of other researchers to serve as a point of reference for future severe accident calculations. The CONTAIN calculations are reviewed to examine several parameters/phenomena of interest. The results of the calculations are also used to identify limitations of the CONTAIN code regarding application to advanced reactor containment designs. The most recent heat transfer correlations available in the literature are assessed for use in the flow regimes and geometries applicable to the AP600. Use of one of these correlations in CONTAIN may allow for a more accurate assessment of the AP600

Experimental study of a water-mist jet issuing normal to a heated flat plate

Directory of Open Access Journals (Sweden)

Vouros Andreas

2016-01-01

Full Text Available A parametric experimental study on the development of a round jet spray impacting a smooth, heated, flat plate has been accomplished. The main objective of this effort was to provide information characterizing the flow structure of a developing mist jet, issuing vertically towards an upward facing, horizontal heated plate, by means of simultaneous droplet size and velocity measurements. Phase Doppler Anemometry was used, providing also information on liquid volume flux. The fine spray of small atomized droplets (0.5-5.0 μm, was generated using a medical nebulizer. Two low Reynolds number jets (Re=2952, 3773 issuing from a cylindrical pipe have been tested. The distance between the jets’ exit and the plate was 50 cm. A stainless steel non-magnetic flat plate of dimensions 1000x500x12mm3 was used as target wall. Constant heat flux boundary conditions were established during measurements. Results indicate that the heat flux from the plate is influencing the evolution of the spray jet, diminishing its velocity and turbulence. Average droplet sizes are affected little by the heat flux, although for the non-heated sprays, droplet sizes increase at locations very close to the plate. A significant effect on droplet volume flow rate is also reported.

Experiments to investigate direct containment heating phenomena with scaled models of the Calvert Cliffs Nuclear Power Plant

International Nuclear Information System (INIS)

Blanchat, T.K.; Pilch, M.M.; Allen, M.D.

1997-02-01

The Surtsey Test Facility is used to perform scaled experiments simulating High Pressure Melt Ejection accidents in a nuclear power plant (NPP). The experiments investigate the effects of direct containment heating (DCH) on the containment load. The results from Zion and Surry experiments can be extrapolated to other Westinghouse plants, but predicted containment loads cannot be generalized to all Combustion Engineering (CE) plants. Five CE plants have melt dispersal flow paths which circumvent the main mitigation of containment compartmentalization in most Westinghouse PWRs. Calvert Cliff-like plant geometries and the impact of codispersed water were addressed as part of the DCH issue resolution. Integral effects tests were performed with a scale model of the Calvert Cliffs NPP inside the Surtsey test vessel. The experiments investigated the effects of codispersal of water, steam, and molten core stimulant materials on DCH loads under prototypic accident conditions and plant configurations. The results indicated that large amounts of coejected water reduced the DCH load by a small amount. Large amounts of debris were dispersed from the cavity to the upper dome (via the annular gap). 22 refs., 84 figs., 30 tabs

Scaling for Mixed Convection Heat Transfer in Passive Containments and Experiment Design

International Nuclear Information System (INIS)

Wang, Shengfei; Yu, Yu; Lv, Xuefeng; Niu, Fenglei; Yan, Xiuping

2012-01-01

Most of the advanced nuclear reactor design utilizes passive systems to remove heat from the core by natural circulation. The passive systems will be widely used in generation III pressurized water reactor. One of the typical passive systems is passive containment cooling system (PCCS), which is a passive condenser system designed to remove heat from the containment for long term cooling after a postulated reactor accident. In order to establish empirical correlations and develop simulation models, a scaling analysis is performed in designing an experiment for the prototype PCCS. This paper presents a scaling method and the design of the experimental facility. The key dimensionless parameters governing the dominant processes are given at last

Comparison of heat stability of goat milk subjected to ultra-high temperature and in-container sterilization.

Science.gov (United States)

Chen, B Y; Grandison, A S; Lewis, M J

2012-03-01

Goat milk with and without stabilizing salt was subjected to in-container and UHT sterilization. Heat stability was assessed by measuring the amount of sediment in the milk. Without stabilizing salts, goat milk usually produced less sediment when subjected to in-container sterilization compared with UHT processing. Addition of stabilizing salts up to 12.8mM resulted in a progressive increase in sediment for in-container sterilization. In contrast, adding stabilizing salts at 6.4mM initially reduced sediment formation in UHT-treated milk but addition of stabilizing salts at 12.8mM increased sediment formation. Adding stabilizing salts to goat milk increased pH, decreased ionic calcium, and increased ethanol stability. Adding up to 2mM calcium chloride increased sediment formation more after UHT treatment than after in-container sterilization. These results suggest that no single mechanism or set of reactions causes milk to produce sediment during heating and that the favored pathway is different for UHT and in-container sterilization processes. Poor heat stability could be induced both by increasing ionic calcium and by decreasing it. Ethanol stability is not a good indicator of heat stability for in-container sterilization, but it may be for UHT sterilization, if milk does not enter the region of poor heat stability found at low concentrations of ionic calcium. Copyright © 2012 American Dairy Science Association. Published by Elsevier Inc. All rights reserved.

Heat, mass, and momentum transport model for hydrogen diffusion flames in nuclear reactor containments

International Nuclear Information System (INIS)

Travis, J.R.

1985-01-01

It is now possible to analyze the time-dependent, fully three-dimensional behavior of hydrogen diffusion flames in nuclear reactor containments. This analysis involves coupling the full Navier-Stokes equations with multi-species transport to the global chemical kinetics of hydrogen combustion. A transport equation for the subgrid scale turbulent kinetic energy density is solved to produce the time and space dependent turbulent transport coefficients. The heat transfer coefficient governing the exchange of heat between fluid computational cells adjacent to wall cells is calculated by a modified Reynolds analogy formulation. The analysis of a MARK-III containment indicates very complex flow patterns that greatly influence fluid and wall temperatures and heat fluxes. 18 refs., 24 figs

GPE-BWR and the containment venting and filtering issue

International Nuclear Information System (INIS)

Palomo, J.; Santiago, J. de

1988-01-01

The Spanish Boiling Water Reactor Owner's Group (GPE-BWR) is formed by three utilities, owning four units: Santa Maria de Garona (46 MWe, BWR3, Mark I containment), Cofrentes (975 MWe, BWR6, Mark III containment) and Valdecaballeros (2x975 MWe, BWR6, Mark III containment) - all of the reactors having been supplied by General Electric. One of the GPE-BWR's several committees is the Safety and Licensing Committee, which follows up the evolution of severe accident topics and particularly the containment venting and filtering issue. In September 1987, the Consejo de Seguridad Nuclear (CSN), the Spanish Regulatory Body, asked the GPE-BWR to define its position on the installation of a containment venting system. The GPE-BWR created a Working Group which presented a Report on Containment Venting to the CSN in January 1987 gathered from: the US Nuclear Regulatory Commission (NRC); some US utilities; and several European countries, especially France, Germany and Sweden. CSN's review of the containment venting Report and the Action Plan proposed by the GPE-BWR finished in April 1988. The conclusion of the Report and the proposed Action Plan take into account the US NRC's identified open items on severe accidents and the R and D programs scheduled to close these items

Oversight of High-Containment Biological Laboratories: Issues for Congress

Science.gov (United States)

2009-05-04

Laboratories: Issues for Congress Congressional Research Service 14 Industry and Non-Profit Laboratories Private sector companies and non-profit...resources for these endeavors. Whether public or private sector , high-containment laboratories are planned and designed to minimize the possibility of... equine encephalitis, and yellow fever. Some of the pathogens that cause these diseases have been considered as biological weapons.104 Expanding the number

Regulatory analysis for the resolution of Generic Issue 99: Loss of RHR [residual heat removal] capability in PWRs

International Nuclear Information System (INIS)

Spano, A.H.

1989-02-01

Generic Issue 99 is concerned with the loss of residual heat removal (RHR) capability in pressurized water reactors during cold-plant outage operations. The issue focuses on two risk-significant common-cause failure modes of the RHR system: (1) air binding of the RHR pumps during reduced-inventory operations and (2) spurious closure of the RHR suction valves due to misapplication of the autoclosure interlocks. Resolution of this issue involves consideration of the adequacy of plant capabilities for (1) preventing losses of RHR, (2) responding promptly and effectively to such challenges in order to prevent core damage, and (3) ensuring timely containment protection against the release of radioactivity to the environment in the unlikely event of core damage due to loss of shutdown cooling. This entails examination of (1) relevant operational and accident response procedures, (2) the instrumentation available to the operator for accident diagnosis and mitigation, and (3) the administrative controls available for ensuring control room cognizance of ongoing maintenance activities that could potentially affect the stability of the reactor coolant system. This regulatory analysis provides quantitative assessments of the costs and benefits associated with several alternatives considered for the resolution of Generic Issue 99. 24 refs

Direct containment heating integral effects tests in geometries of European nuclear power plants

International Nuclear Information System (INIS)

Meyer, Leonhard; Albrecht, Giancarlo; Caroli, Cataldo; Ivanov, Ivan

2009-01-01

The DISCO test facility at Forschungszentrum Karlsruhe (FZK) has been used to perform experiments to investigate direct containment heating (DCH) effects during a severe accident in European nuclear power plants, comprising the EPR, the French 1300 MWe plant P'4, the VVER-1000 and the German Konvoi plant. A high-temperature iron-alumina melt is ejected by steam into scaled models of the respective reactor cavities and the containment vessel. Both heat transfer from dispersed melt and combustion of hydrogen lead to containment pressurization. The main experimental findings are presented and critical parameters are identified. The consequences of DCH are limited in reactors with no direct pathway between the cavity and the containment dome (closed pit). The situation is more severe for reactors which do have a direct pathway between the cavity and the containment (open pit). The experiments showed that substantial fractions of corium may be dispersed into the containment in such cases, if the pressure in the reactor coolant system is elevated at the time of RPV failure. Primary system pressures of 1 or 2 MPa are sufficient to lead to full scale DCH effects. Combustion of the hydrogen produced by oxidation as well as the hydrogen initially present appears to be the crucial phenomenon for containment pressurization.

Direct containment heating integral effects tests in geometries of European nuclear power plants

Energy Technology Data Exchange (ETDEWEB)

Meyer, Leonhard [Forschungszentrum Karlsruhe (FZK), Postfach 3640, 76021 Karlsruhe (Germany)], E-mail: meyer@iket.fzk.de; Albrecht, Giancarlo [Forschungszentrum Karlsruhe (FZK), Postfach 3640, 76021 Karlsruhe (Germany); Caroli, Cataldo [Institut de Radioprotection et de Surete Nucleaire, BP 17, 92262 Fontenay-aux-Roses Cedex (France); Ivanov, Ivan [Technical University of Sofia, BG-1797 Sofia (Bulgaria)

2009-10-15

The DISCO test facility at Forschungszentrum Karlsruhe (FZK) has been used to perform experiments to investigate direct containment heating (DCH) effects during a severe accident in European nuclear power plants, comprising the EPR, the French 1300 MWe plant P'4, the VVER-1000 and the German Konvoi plant. A high-temperature iron-alumina melt is ejected by steam into scaled models of the respective reactor cavities and the containment vessel. Both heat transfer from dispersed melt and combustion of hydrogen lead to containment pressurization. The main experimental findings are presented and critical parameters are identified. The consequences of DCH are limited in reactors with no direct pathway between the cavity and the containment dome (closed pit). The situation is more severe for reactors which do have a direct pathway between the cavity and the containment (open pit). The experiments showed that substantial fractions of corium may be dispersed into the containment in such cases, if the pressure in the reactor coolant system is elevated at the time of RPV failure. Primary system pressures of 1 or 2 MPa are sufficient to lead to full scale DCH effects. Combustion of the hydrogen produced by oxidation as well as the hydrogen initially present appears to be the crucial phenomenon for containment pressurization.

Mathematical modelling of heat absorption capacity of containment spray system in a 700 MWe PHWR

International Nuclear Information System (INIS)

Kota, Sampath Bharadwaj; Ali, Seik Mansoor; Balasubramaniyan, V.

2015-01-01

This paper presents a mathematical model for estimating the heat removal by containment spray system in the post Loss of Coolant Accident (LOCA) environment. The procedure involves firstly, the calculation of heat removal rates by droplets of spray dispersed in the air-steam mixture by an appropriate direct contact condensation model accounting for the presence of non-condensable gas (air). Parametric influence of droplet size, ambient pressure and temperature on heat flux is brought out. It was found that the heat flux is inversely proportional to the ambient pressure and diameter. A spray module was subsequently developed and incorporated into an in-house containment thermal hydraulics code. The pressure and temperature transients in a 700 MWe PHWR containment building following a Large Break LOCA was obtained using this code. The efficacy of the spray in condensing the steam is shown by comparing the transients with and without the operation of spray system. Parametric studies are also conducted with respect to droplet size and flow rate of water droplet spray. The details of the investigation are presented and discussed in this paper. (author)

Safety issues related to the intermediate heat storage for the EU DEMO

Energy Technology Data Exchange (ETDEWEB)

Carpignano, Andrea [NEMO group, Dipartimento Energia, Politecnico di Torino, C.so Duca degli Abruzzi 24, 10129 Torino (Italy); Pinna, Tonio [ENEA, 00044 Frascati (Italy); Savoldi, Laura; Sobrero, Giulia; Uggenti, Anna Chiara [NEMO group, Dipartimento Energia, Politecnico di Torino, C.so Duca degli Abruzzi 24, 10129 Torino (Italy); Zanino, Roberto, E-mail: roberto.zanino@polito.it [NEMO group, Dipartimento Energia, Politecnico di Torino, C.so Duca degli Abruzzi 24, 10129 Torino (Italy)

2016-11-01

Highlights: • IHS affects only the PHTS and the BoP (Balance of Plant). • PIEs list does not change but IHS influences PIEs evolution. • Additional issues to be addressed in PIEs study due to the implementation of HIS. • No safety/operational major obstacles were found for IHS concept. - Abstract: The functional deviations able to compromise system safety in the EU DEMO Primary Heat Transfer System (PHTS) with intermediate heat storage (IHS) based on molten salts are identified and compared to the deviations identified with PHTS without IHS. The resulting safety issues for the Balance of Plant (BoP) have been taken into account. Functional Failure Mode and Effects Analysis (FFMEA) is used to highlight the Postulated Initiating Events (PIE) of incident/accident sequences and to provide some safety insights during the preliminary design. The architecture of the system with IHS does not introduce new PIE with respect to the case without IHS, but it modifies some of them. In particular the two Postulated Initiating Events that are affected by the presence of IHS are the LOCA in the tubes of the HX between primary and intermediate circuit and the loss of heat sink for the first wall or the breeding zone. In fact the IHS introduces some advantages concerning the stability of the secondary circuit, but some weaknesses are associated to the physical-chemical nature of molten salts, especially oxidizing power, corrosive nature and risk of solidification. These issues can be managed in the design by the introduction of new safety functions.

29 CFR 1915.54 - Welding, cutting and heating of hollow metal containers and structures not covered by § 1915.12.

Science.gov (United States)

2010-07-01

... 29 Labor 7 2010-07-01 2010-07-01 false Welding, cutting and heating of hollow metal containers and... STANDARDS FOR SHIPYARD EMPLOYMENT Welding, Cutting and Heating § 1915.54 Welding, cutting and heating of... which have contained flammable substances shall, before welding, cutting, or heating is undertaken on...

COMMIX analysis of AP-600 Passive Containment Cooling System

International Nuclear Information System (INIS)

Chang, J.F.C.; Chien, T.H.; Ding, J.; Sun, J.G.; Sha, W.T.

1992-01-01

COMMIX modeling and basic concepts that relate components, i.e., containment, water film cooling, and natural draft air flow systems. of the AP-600 Passive Containment Cooling System are discussed. The critical safety issues during a postulated accident have been identified as (1) maintaining the liquid film outside the steel containment vessel, (2) ensuring the natural convection in the air annulus. and (3) quantifying both heat and mass transfer accurately for the system. The lack of appropriate heat and mass transfer models in the present analysis is addressed. and additional assessment and validation of the proposed models is proposed

Heat transfer analysis of the waste-container sleeve/salt configuration

International Nuclear Information System (INIS)

Callahan, G.D.; Ratigan, J.L.; Russell, J.E.; Fossum, A.F.

1975-01-01

Prior to this investigation, the heat transport considered was only that of straight conduction. The waste container, air gap, and sleeve arrangement was considered to be a single, consistent, time-dependent, heat-generating unit in intimate contact with the salt. The conduction model does not accurately model the heat transfer mechanisms available. Thus radiation and combined radiation and convection must also be considered in the determination of the temperature field. As would be expected, the canister temperatures are higher for the case of radiation across the airgap than those that result from conduction when the canister is in intimate contact with the salt. For the radiation case, the canister temperatures rise rapidly to a temperature of approximately 1,140 0 F and maintain an almost steady state condition for one year whereafter the temperatures slowly decrease. The far field temperatures, near the pillar centerline, are essentially equivalent for all cases. As time proceeds, the far field temperatures of the conduction models are about 15% different

Containment-emergency-sump performance. Technical findings related to Unresolved Safety Issue A-43

International Nuclear Information System (INIS)

1983-04-01

This report summarizes key technical findings related to the Unresolved Safety Issue A-43, Containment Emergency Sump Performance, and provides recommendations for resolution of attendant safety issues. The key safety questions relate to: (a) effects of insulation debris on sump performance; (b) sump hydraulic performance as determined by design features, submergence, and plant induced effects, and (c) recirculation pump performance wherein air and/or particulate ingestion can occur. The technical findings presented in this report provide information relevant to the design and performance evaluation of the containment emergency sump

Analysis of the passive heat removal enhancement for AP1000 containment due to the partially wetted coverage

Energy Technology Data Exchange (ETDEWEB)

Li, Cheng, E-mail: 510395453@qq.com [State Nuclear Power Technology Research & Development Center, 102209 Beijing (China); Li, Le [Tsinghua University, Institute of Nuclear and New Energy Technology, 100084 Beijing (China); Li, Junming [Tsinghua University, Key Laboratory for Thermal Science and Power Engineering of Ministry of Education, Department of Thermal Engineering, Beijing 100084 (China); Zhang, Yajun [Tsinghua University, Institute of Nuclear and New Energy Technology, 100084 Beijing (China); Li, Zhihui [State Nuclear Power Technology Research & Development Center, 102209 Beijing (China)

2017-03-15

Highlights: • Heat removal by steam condensation, thermal conduction and evaporation is the most important scheme for AP1000 PCCS. Traditionally, studies on containment wall condensation and evaporation have been widely made, while it lacks studies on the shell two-dimension (2-D) thermal conduction. Currently, based on the known heat and mass transfer correlations and the phenomenon from water wetted coverage test, the physical model for 2-D thermal conduction is given and numerical simulation is then made. By discussions, it forms the following highlights. • The partially wetted surface can enhance the whole heat transfer process (including inner condensation, wall thermal conduction and outside cooling) and the maximum enhancement factor can be as large as 63%. There is an enhancement peak at around dry strip fraction a = 90%. When L is less than 0.03 m, its influence on heat transfer is small and the enhancement is mainly affected by dry coverage. However, for larger L, both α and L contribute much to larger enhancement. • Location at the spring line is often used for safety analysis and the dry strip fraction there for AP1000 is mainly at 10%–80%. Accordingly, further analysis is made on L (0.03 < L < 0.3) and a fitting expression is given for α = 10%–80%. It could be used to improve the corresponding software and it could also be used for containment scaling-down criteria analysis. - Abstract: AP1000 containment uses the water film evaporation, coupled with containment inner condensation, to remove the core decay heat. However, water film cannot fully cover heat transfer surface and dry-wetted strips appear. As a result, heat transfer within the containment shell is a two-dimension thermal conduction. Current work numerically studied the AP1000 heat removal enhancement due to the partially wetted coverage phenomenon. It used the evaporation and condensation boundary conditions and Fluent software to calculate the local heat fluxes and their

Analysis of the passive heat removal enhancement for AP1000 containment due to the partially wetted coverage

International Nuclear Information System (INIS)

Li, Cheng; Li, Le; Li, Junming; Zhang, Yajun; Li, Zhihui

2017-01-01

Highlights: • Heat removal by steam condensation, thermal conduction and evaporation is the most important scheme for AP1000 PCCS. Traditionally, studies on containment wall condensation and evaporation have been widely made, while it lacks studies on the shell two-dimension (2-D) thermal conduction. Currently, based on the known heat and mass transfer correlations and the phenomenon from water wetted coverage test, the physical model for 2-D thermal conduction is given and numerical simulation is then made. By discussions, it forms the following highlights. • The partially wetted surface can enhance the whole heat transfer process (including inner condensation, wall thermal conduction and outside cooling) and the maximum enhancement factor can be as large as 63%. There is an enhancement peak at around dry strip fraction a = 90%. When L is less than 0.03 m, its influence on heat transfer is small and the enhancement is mainly affected by dry coverage. However, for larger L, both α and L contribute much to larger enhancement. • Location at the spring line is often used for safety analysis and the dry strip fraction there for AP1000 is mainly at 10%–80%. Accordingly, further analysis is made on L (0.03 < L < 0.3) and a fitting expression is given for α = 10%–80%. It could be used to improve the corresponding software and it could also be used for containment scaling-down criteria analysis. - Abstract: AP1000 containment uses the water film evaporation, coupled with containment inner condensation, to remove the core decay heat. However, water film cannot fully cover heat transfer surface and dry-wetted strips appear. As a result, heat transfer within the containment shell is a two-dimension thermal conduction. Current work numerically studied the AP1000 heat removal enhancement due to the partially wetted coverage phenomenon. It used the evaporation and condensation boundary conditions and Fluent software to calculate the local heat fluxes and their

Use of an integrated containment and ultimate heat sink (UHS) response approach to evaluate nuclear power plant modifications

International Nuclear Information System (INIS)

Wetzel, M.C.; Vieira, A.T.; Patton, D.C.

1994-01-01

Detailed containment and Ultimate Heat Sink (UHS) performance evaluations often are required to support major plant modifications, such as power up-rates and steam generator replacements. These UHS and containment pressure and temperature response evaluations are interrelated. Not only is the containment heat load to the UHS a factor in these evaluations, but other heat loads, such as those from the spent fuel pool, may change as a result of the plant modification and impact containment or UHS response. Our experience is that if an integrated containment/UHS response model is developed prior to the feasibility evaluations for such plant modifications, significant savings in engineering hours can be achieved. This paper presents an overview of such a front-end engineering tool that has been developed and used to support engineering evaluations. 3 refs., 2 figs

Boiling Heat Transfer Coefficients of Nanofluids Containing Carbon Nanotubes up to Critical Heat Fluxes

International Nuclear Information System (INIS)

Park, Ki Jung; Lee, Yohan; Jung, Dong Soo; Shim, Sang Eun

2011-01-01

In this study, the nucleate pool boiling heat transfer coefficients (HTCs) and critical heat flux (CHF) for a smooth and square flat heater in a pool of pure water with and without carbon nanotubes (CNTs) dispersed at 60 .deg. C were measured. Tested aqueous nanofluids were prepared using CNTs with volume concentrations of 0.0001%, 0.001%, and 0.01%. The CNTs were dispersed by chemically treating them with an acid in the absence of any polymers. The results showed that the pool boiling HTCs of the nanofluids are higher than those of pure water in the entire nucleate boiling regime. The acid-treated CNTs led to the deposition of a small amount of CNTs on the surface, and the CNTs themselves acted as heat-transfer-enhancing particles, owing to their very high thermal conductivity. There was a significant increase in the CHF- up to 150%-when compared to that of pure water containing CNTs with a volume concentration of 0.001%. This is attributed to the change in surface characteristics due to the deposition of a very thin layer of CNTs on the surface. This layer delays nucleate boiling and causes a reduction in the size of the large vapor canopy around the CHF. This results in a significant increase in the CHF

Direct waste heat recovery via thermoelectric materials - chosen issues of the thermodynamic description

International Nuclear Information System (INIS)

Kolasiński, Piotr; Kolasińska, Ewa

2016-01-01

The effective waste heat recovery is one of the present-day challenges in the industry and power engineering. The energy systems dedicated for waste heat conversion into electricity are usually characterized by low efficiency and are complicated in the design. The possibility of waste heat recovery via thermoelectric materials may be an interesting alternative to the currently used technologies. In particular, due to their material characteristics, conducting polymers may be competitive when compared with the power machinery and equipment. These materials can be used in a wide range of the geometries e.g. the bulk products, thin films, pristine form or composites and the others. In this article, the authors present selected issues related to the mathematical and thermodynamic description of the heat transfer processes in the thermoelectric materials dedicated for the waste heat recovery. The link of these models with electrical properties of the material and a material solution based on a conducting polymer have also been presented in this paper. (paper)

Problems of heat transfer within the containing vessel of high performance LMFBR spent fuel shipping casks

International Nuclear Information System (INIS)

Pope, R.B.; Gartling, D.K.; Schimmel, W.P. Jr.; Larson, D.W.

1976-01-01

A preliminary assessment of heat transfer problems internal to a LMFBR spent fuel shipping cask is reported. The assessment is based upon previous results obtained in full-scale, electrically heated mockups of an LMFBR assembly located in a containing pipe, and also upon analytical and empirical studies presented in this paper. It is shown that a liquid coolant will be required to adequately distribute the decay heat of short-cooled assemblies from the fuel region to the containing cask structure. Liquid sodium apparently provides the best heat transfer, and sufficient data are available to adequately model the heat transfer processes involved. Dowtherm A is the most efficient organic evaluated to date and presented in the open literature. Since the organic materials have high Prandtl and usually high Rayleigh numbers, natural convection is the predominant mode of heat transfer. It is shown that a more comprehensive understanding of the convective processes will be required before heat transfer with an organic coolant can be adequately modeled. However, in view of systems considerations, Dowtherm A should be further considered as an alternative to sodium for use as a LMFBR spent fuel shipping cask coolant

Evaporation and condensation heat transfer in a suppression chamber of the water wall type passive containment cooling system

International Nuclear Information System (INIS)

Fujii, Tadashi; Kataoka, Yoshiyuki; Murase, Michio

1996-01-01

To evaluate the system pressure response of a water wall type containment cooling system, which is one of the passive safety systems, the evaporation and condensation behaviors in a suppression chamber have been experimentally examined. In the system, the suppression pool water evaporates from the pool surface, passing into the wetwell due to pool temperature rise, while steam in the wetwell condenses on the steel containment vessel wall due to the heat release through the wall. The wetwell is a gas phase region in the suppression chamber and its pressure, which is expressed as the sum of the noncondensable gas pressure and saturated steam pressure, is strongly affected by the evaporation heat transfer from the suppression pool surface and condensation heat transfer on the containment vessel wall. Based on the measured temperature profiles near the heat transfer surface and the wetwell pressure using two apparatuses, evaporation and condensation heat transfer coefficients were evaluated. The following results were obtained. (1) Both heat transfer coefficients increased as the ratio of the steam partial pressure to the total pressure increased. (2) Comparison of the results from two types of test apparatuses confirmed that the size of the heat transfer surface did not affect the heat transfer characteristics within these tests. (3) The heat transfer coefficients were expressed by the ratio of the steam to noncondensable gas logarithmic mean concentration, which considered the steam and gas concentration gradient from the heat transfer surface to the wetwell bulk. (author)

Nuclear reactor melt-retention structure to mitigate direct containment heating

International Nuclear Information System (INIS)

Tutu, N.K.; Ginsberg, T.; Klages, J.R.

1991-01-01

This patent describes a nuclear reactor melt-retention structure that functions to retain molten core material within a melt retention chamber to mitigate the extent of direct containment heating. The structure being adapted to be positioned within or adjacent to a pressurized or boiling water nuclear reactor containment building at a location such that at least a portion of the melt retention structure is lower than and to one side of the nuclear reactor pressure vessel, and such that the structure is adjacent to a gas escape channel means that communicates between the reactor cavity and the containment building of the reactor. It comprises a melt-retention chamber, wall means defining a passageway extending between the reactor cavity underneath the reactor pressure vessel and one side of the chamber, the passageway including vent means extending through an upper wall portion thereof. The vent means being in communication with the upper region of the reactor containment building, whereby gas and steam discharged from the reactor pressure vessel are vented through the passageway and vent means into the gas-escape channel means and the reactor containment building

Generating debate on issues surrounding the venting of containment in the event of LOCA to ensure an optimised safe outcome

International Nuclear Information System (INIS)

Chadwick, Chris; Jahouel, Xavier; Swain, Adam

2014-01-01

Following the incidents at the Japanese Fukushima Nuclear facility, when three units experienced LOCA, the consequences of those events have caused ripples across the world. Regulators around the world are examining the need to prevent excessive pressurisation of NPP Containment, and safely evacuate the gaseous consequences of LOCA, there is a need to return to fundamental principles and examine each putative set of data derived from the various models describing the consequence of LOCA and other events, a LOCA being a 'Loss Of Coolant Accident', and represents the outcome from a range of incidents ranging from fuel pellets being exposed to cooling (heat exchange) water, to a complete melt down of the fuel load with consequence evaporation of the concrete structures of the reactor containment buildings. Clearly it would be highly desirable in both economic and safety terms to minimise the external impact of these events inside the containment. Since one of the results of a LOCA is the pressurisation of the internal space of the containment building, regulators and the industry are looking at the mandatory installation of suitable vent systems to vent the pressure building up inside the containment, whilst ensuring the minimum impact on the surrounding environment. This is clearly a filtration/separation/recombination issue, and as an expert engineering company in the nuclear industry, Porvair has concerns that the issues of Containment Venting are not being addressed by expert filter companies, but by expert nuclear engineering companies with only a superficial knowledge of the complexities and nuances on filtration processes. The paper will describe in depth and detail the individual consequences of each particular aspect of the modelled data. Looking at flowing conditions (pressure, temperature, gas constituents including water vapour and Caesium and Iodine compounds), vent pressure philosophy, deposition of solids (size, type and quantity), decay heat

Alkali Metal Heat Pipe Life Issues

International Nuclear Information System (INIS)

Reid, Robert S.

2004-01-01

One approach to fission power system design uses alkali metal heat pipes for the core primary heat-transfer system. Heat pipes may also be used as radiator elements or auxiliary thermal control elements. This synopsis characterizes long-life core heat pipes. References are included where information that is more detailed can be found. Specifics shown here are for demonstration purposes and do not necessarily reflect current Nasa Project Prometheus point designs. (author)

The probability of containment failure by direct containment heating in Zion. Supplement 1

International Nuclear Information System (INIS)

Pilch, M.M.; Allen, M.D.; Stamps, D.W.; Tadios, E.L.; Knudson, D.L.

1994-12-01

Supplement 1 of NUREG/CR-6075 brings to closure the DCH issue for the Zion plant. It includes the documentation of the peer review process for NUREG/CR-6075, the assessments of four new splinter scenarios defined in working group meetings, and modeling enhancements recommended by the working groups. In the four new scenarios, consistency of the initial conditions has been implemented by using insights from systems-level codes. SCDAP/RELAP5 was used to analyze three short-term station blackout cases with Different lead rates. In all three case, the hot leg or surge line failed well before the lower head and thus the primary system depressurized to a point where DCH was no longer considered a threat. However, these calculations were continued to lower head failure in order to gain insights that were useful in establishing the initial and boundary conditions. The most useful insights are that the RCS pressure is-low at vessel breach metallic blockages in the core region do not melt and relocate into the lower plenum, and melting of upper plenum steel is correlated with hot leg failure. THE SCDAP/RELAP output was used as input to CONTAIN to assess the containment conditions at vessel breach. The containment-side conditions predicted by CONTAIN are similar to those originally specified in NUREG/CR-6075

Heat exchanger, particularly liquid sodium heated steam generator

International Nuclear Information System (INIS)

Robin, Marcel; Tillequin, Jean.

1977-01-01

This invention relates to a liquid sodium heated steam generator the characteristic of which is an annular distribution chamber fed by two independent and diametrically opposed manifolds on a common horizontal axis, issuing respectively into two adjacent compartments made in the chambers on both sides of a vertical transversal partition containing the axis of the casing and extending perpendicularly to the manifolds, each compartment being itself divided into a number of adjacent sectors marked by folded metal sheets fixed to the distributor and shaped so as to present in pairs and with the chamber opposite the manifold issuing into a compartment two independent ducts for distributing the sodium flow [fr

Heat cascading regenerative sorption heat pump

Science.gov (United States)

Jones, Jack A. (Inventor)

1995-01-01

A simple heat cascading regenerative sorption heat pump process with rejected or waste heat from a higher temperature chemisorption circuit (HTCC) powering a lower temperature physisorption circuit (LTPC) which provides a 30% total improvement over simple regenerative physisorption compression heat pumps when ammonia is both the chemisorbate and physisorbate, and a total improvement of 50% or more for LTPC having two pressure stages. The HTCC contains ammonia and a chemisorbent therefor contained in a plurality of canisters, a condenser-evaporator-radiator system, and a heater, operatively connected together. The LTPC contains ammonia and a physisorbent therefor contained in a plurality of compressors, a condenser-evaporator-radiator system, operatively connected together. A closed heat transfer circuit (CHTC) is provided which contains a flowing heat transfer liquid (FHTL) in thermal communication with each canister and each compressor for cascading heat from the HTCC to the LTPC. Heat is regenerated within the LTPC by transferring heat from one compressor to another. In one embodiment the regeneration is performed by another CHTC containing another FHTL in thermal communication with each compressor. In another embodiment the HTCC powers a lower temperature ammonia water absorption circuit (LTAWAC) which contains a generator-absorber system containing the absorbent, and a condenser-evaporator-radiator system, operatively connected together. The absorbent is water or an absorbent aqueous solution. A CHTC is provided which contains a FHTL in thermal communication with the generator for cascading heat from the HTCC to the LTAWAC. Heat is regenerated within the LTAWAC by transferring heat from the generator to the absorber. The chemical composition of the chemisorbent is different than the chemical composition of the physisorbent, and the absorbent. The chemical composition of the FHTL is different than the chemisorbent, the physisorbent, the absorbent, and ammonia.

Technological issues of ion cyclotron heating of fusion plasmas

International Nuclear Information System (INIS)

Hwang, D.Q.; Fortgang, C.M.

1985-01-01

With the recent promising results of plasma heating using electromagnetic waves (EM waves) in the ion cyclotron range of frequency (ICRF) on the Princeton Large Torus (PLT) tokamak the feasibility of employing ICRF heating to a reactor-like magnetic confinement device is increasing. The high power ICRF experiments funded on JET (Joint European Torus in England) and JT-60 (in Japan) will have rf source power in the range of 10-30 MW. The time scale for the duration of the RF pulse will range from seconds up to steady-state. The development of new RF components that can transmit and launch such high power, long pulse length, EM waves in a plasma environment is a major technological task. In general, the technology issues may be divided into two categories. The first category concerns the region where the plasma comes in contact with the wave launchers. The problems here are dominated by plasmamaterial interaction, heat deposition by the plasma onto the wave launcher, and erosion of the launcher material. It is necessary to minimize the heat deposition from the plasma, the losses of the RF wave energy in the structure, and to prevent sputtering of the antenna components. A solution involves a combined design using special materials and optimal shaping of the Faraday shield (the electrostatic shields which can be used both for an EM wave polarization adjustment and as a particle shield for the launcher). Recent studies by PPPL and McDonnell Douglas Corp. on the Faraday shield designs will be discussed. The second important area where technology development will be necessary is the transmission of high power RF waves through a gas/vacuum interface region. In the past, the vacuum feedthrough has been the bottle neck which prevented high power operation of the PLT antenna

Weak Solution and Weakly Uniformly Bounded Solution of Impulsive Heat Equations Containing “Maximum” Temperature

Directory of Open Access Journals (Sweden)

Oyelami, Benjamin Oyediran

2013-09-01

Full Text Available In this paper, criteria for the existence of weak solutions and uniformly weak bounded solution of impulsive heat equation containing maximum temperature are investigated and results obtained. An example is given for heat flow system with impulsive temperature using maximum temperature simulator and criteria for the uniformly weak bounded of solutions of the system are obtained.

Containment integrity analysis with SAMPSON/DCRA module

International Nuclear Information System (INIS)

Hosoda, Seigo; Shirakawa, Noriyuki; Naitoh, Masanori

2006-01-01

The integrity of PWR containment under a severe accident is analyzed using the debris concrete reaction analysis code. If core fuels melt through the pressure vessel and the debris accumulates on the reactor cavity of a lower part of containment, its temperature continues to rise due to decay heat and the debris ablates the concrete floor. In case that cooling water is issued into the containment cavity and the amount of debris is limited to 30% of core fuels, our analyses showed that the debris could be cooled and frozen so that integrity of containment could hold. (author)

Hydride heat pump with heat regenerator

Science.gov (United States)

Jones, Jack A. (Inventor)

1991-01-01

A regenerative hydride heat pump process and system is provided which can regenerate a high percentage of the sensible heat of the system. A series of at least four canisters containing a lower temperature performing hydride and a series of at least four canisters containing a higher temperature performing hydride is provided. Each canister contains a heat conductive passageway through which a heat transfer fluid is circulated so that sensible heat is regenerated. The process and system are useful for air conditioning rooms, providing room heat in the winter or for hot water heating throughout the year, and, in general, for pumping heat from a lower temperature to a higher temperature.

Microencapsulated Phase-Change Materials For Storage Of Heat

Science.gov (United States)

Colvin, David P.

1989-01-01

Report describes research on engineering issues related to storage and transport of heat in slurries containing phase-change materials in microscopic capsules. Specific goal of project to develop lightweight, compact, heat-management systems used safely in inhabited areas of spacecraft. Further development of obvious potential of technology expected to lead to commercialization and use in aircraft, electronic equipment, machinery, industrial processes, and other sytems in which requirements for management of heat compete with severe restrictions on weight or volume.

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

Tooth bleaching using peroxide-containing agents: current status of safety issues.

Science.gov (United States)

Li, Y

1998-08-01

During the last 10 years, at-home tooth bleaching using peroxide-containing agents has quickly become well accepted. It is one of the most popular cosmetic dental procedures for whitening teeth. Although there are few disputes regarding their efficacy, concerns and debates have continued regarding the safety of peroxide-containing tooth bleaching agents. Potential carcinogenicity and genotoxicity of the peroxides used in bleaching agents are the two most persistent and controversial issues. This article reviews and discusses available information on carcinogenicity and genotoxicity of hydrogen peroxide and the potential risks associated with the use of peroxide-containing bleaching agents. Clinical studies reported after the 1996 international symposium on responses of oral tissues to peroxide-containing bleaching agents are also reviewed. Overall evidence supports the conclusion that the proper use of peroxide-containing at-home tooth bleaching agents is safe. However, potential adverse effects may occur in inappropriate applications, abuses, or the use of inappropriate products. At-home tooth bleaching should be monitored by dental professionals to maximize the benefits while minimizing potential risks.

Impact of bulk atmospheric motion on local and global containment heat transfer

International Nuclear Information System (INIS)

Green, J.A.; Almenas, K.

1995-01-01

Local and global correlations for condensing energy transfer in the presence of noncondensable gases in a containment facility have been evaluated. The database employed stems from the E11.2 and E11.4 tests conducted at the German HDR facility. The HDR containment is a 11060-ml, 60-m-high decommissioned light water reactor. The tests simulated long-term (up to 56 h) accident conditions. Numerous instrumented structural blocks (concrete and lead) were located throughout the containment to provide detailed local heat transfer measurements. These data represent what is probably the most extensive database of integral energy transfer measurements available. It is well established that the major resistance to condensation heat transfer in the presence of noncondensable gases is a gaseous boundary layer that builds up in front of the condensing surface. Correlations that seek to model heat transfer for these conditions should depend on parameters that most strongly determine the buildup and thickness of this boundary layer. Two of the most important parameters are the vapor/noncondensable concentration ratio and the local atmospheric motion. Secondary parameters include the atmosphere-to-surface temperature difference, the pressure, and condensing surface properties. The HDR tests are unique in terms of the quantity and variety of instrumentation employed. However, one of the most important parameters, the local bulk atmospheric velocity, is inherently difficult to measure, and only fragmentary measurements are available even in the HDR data-base. A detailed analysis of these data is presented by Green. This study uses statistical methods to evaluate local and global empirical correlations that do not include the atmospheric velocity. The magnitude of the differences between the correlations emphasizes the importance of the local atmospheric velocity and serves to illustrate the accuracy limits of correlations that neglect this essential parameter

Aerodynamics, heat and mass transfer in steam-aerosol turbulent flows in containment

Energy Technology Data Exchange (ETDEWEB)

Nigmatulin, B.I.; Pershukov, V.A.; Ris, V.V. [Research & Engineering Centre of Nuclear Plants Safety, Moscow (Russian Federation)] [and others

1995-09-01

In this report an analysis of aerodynamic and heat transfer processes at the blowdown of gas-dispersed mixture into the containment volume is presented. A few models for description of the volume averaged and local characteristics are analyzed. The mathematical model for description of the local characteristics of the turbulent gas-dispersed flows was developed. The calculation of aerodynamic, heat and mass transfer characteristics was based on the Navier-Stokes, energy and gas mass fractions conservation equations. For calculation of dynamics and deposition of the aerosols the original diffusion-inertia model is developed. The pulsating characteristics of the gaseous phase were calculated on the base (k-{xi}) model of turbulence with modification to account thermogravitational force action and influence of particle mass loading. The appropriate boundary conditions using the {open_quotes}near-wall function{close_quotes} approach was obtained. Testing of the mathematical models and boundary conditions has shown a good agreement between computation and data of comparison. The described mathematical models were applied to two- and three dimensional calculations of the turbulent flow in containment at the various stages of the accident.

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

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.

Vitrification of high level nuclear waste inside ambient temperature disposal containers using inductive heating: The SMILE system

International Nuclear Information System (INIS)

Powell, J.; Reich, M.; Barletta, R.

1996-01-01

A new approach, termed SMILE (Small Module Inductively Loaded Energy), for the vitrification of high level nuclear wastes (HLW) is described. Present vitrification systems liquefy the HLW solids and associated frit material in large high temperature melters. The molten mix is then poured into small (∼1 m 3 ) disposal canisters, where it solidifies and cools. SMILE eliminates the separate, large high temperature melter. Instead, the BLW solids and frit melt inside the final disposal containers, using inductive heating. The contents then solidify and cool in place. The SMILE modules and the inductive heating process are designed so that the outer stainless can of the module remains at near ambient temperature during the process cycle. Module dimensions are similar to those of present disposal containers. The can is thermally insulated from the high temperature inner container by a thin layer of refractory alumina firebricks. The inner container is a graphite crucible lined with a dense alumina refractory that holds the HLW and fiit materials. After the SMILE module is loaded with a slurry of HLW and frit solids, an external multi-turn coil is energized with 30-cycle AC current. The enclosing external coil is the primary of a power transformer, with the graphite crucible acting as a single turn ''secondary.'' The induced current in the ''secondary'' heats the graphite, which in turn heats the HLW and frit materials. The first stage of the heating process is carried out at an intermediate temperature to drive off remnant liquid water and water of hydration, which takes about 1 day. The small fill/vent tube to the module is then sealed off and the interior temperature raised to the vitrification range, i.e., ∼1200C. Liquefaction is complete after approximately 1 day. The inductive heating then ceases and the module slowly loses heat to the environment, allowing the molten material to solidify and cool down to ambient temperature

Nuclear reactor melt-retention structure to mitigate direct containment heating

Science.gov (United States)

Tutu, Narinder K.; Ginsberg, Theodore; Klages, John R.

1991-01-01

A light water nuclear reactor melt-retention structure to mitigate the extent of direct containment heating of the reactor containment building. The structure includes a retention chamber for retaining molten core material away from the upper regions of the reactor containment building when a severe accident causes the bottom of the pressure vessel of the reactor to fail and discharge such molten material under high pressure through the reactor cavity into the retention chamber. In combination with the melt-retention chamber there is provided a passageway that includes molten core droplet deflector vanes and has gas vent means in its upper surface, which means are operable to deflect molten core droplets into the retention chamber while allowing high pressure steam and gases to be vented into the upper regions of the containment building. A plurality of platforms are mounted within the passageway and the melt-retention structure to direct the flow of molten core material and help retain it within the melt-retention chamber. In addition, ribs are mounted at spaced positions on the floor of the melt-retention chamber, and grid means are positioned at the entrance side of the retention chamber. The grid means develop gas back pressure that helps separate the molten core droplets from discharged high pressure steam and gases, thereby forcing the steam and gases to vent into the upper regions of the reactor containment building.

Radiant heat testing of the H1224A shipping/storage container

Energy Technology Data Exchange (ETDEWEB)

Harding, D.C.; Bobbe, J.G.; Stenberg, D.R.; Arviso, M.

1994-05-01

H1224A weapons containers have been used for years by the Departments of Energy and Defense to transport and store W78 warhead midsections. Although designed to protect the midsections only from low-energy impacts, a recent transportation risk assessment effort has identified a need to evaluate the container`s ability to protect weapons in more severe accident environments. Four radiant heat tests were performed: two each on an H1224A container (with a Mk12a Mod 6c mass mock-up midsection inside) and two on a low-cost simulated H1224A container (with a hollow Mk12 aeroshell midsections inside). For each unit tested, temperatures were recorded at numerous points throughout the container and midsection during a 4-hour 121{degrees}C (250{degrees}F) and 30-minute 1010{degrees}C (1850{degrees}F) radiant environment. Measured peak temperatures experienced by the inner walls of the midsections as a result of exposure to the high-temperature radiant environment ranged from 650{degrees} C to 980{degrees} C (1200{degrees} F to 1800{degrees}F) for the H1224A container and 770 {degrees} to 990 {degrees}C (1420{degrees} F to 1810{degrees}F) for the simulated container. The majority of both containers were completely destroyed during the high-temperature test. Temperature profiles will be used to benchmark analytical models and predict warhead midsection temperatures over a wide range of the thermal accident conditions.

Containment fan cooler heat transfer calculation during main steam line break for Maanshan PWR plant

Energy Technology Data Exchange (ETDEWEB)

Yuann, Yng-Ruey, E-mail: ryyuann@iner.gov.tw; Kao, Lain-Su, E-mail: lskao@iner.gov.tw

2013-10-15

Highlights: • Evaluate component cooling water (CCW) thermal response during MSLB for Maanshan. • Using GOTHIC to calculate CCW temperature and determine time required to boil CCW. • Both convective and condensation heat transfer from the air side are considered. • Boiling will not occur since T{sub B} is sufficiently longer than CCW pump restart time. -- Abstract: A thermal analysis has been performed for the Containment Fan Cooler Unit (FCU) during Main Steam Line Break (MSLB) accident, concurrent with loss of offsite power, for Maanshan PWR plant. The analysis is performed in order to address the waterhammer and two-phase flow issues discussed in USNRC's Generic Letter 96-06 (GL 96-06). Maanshan plant is a twin-unit Westinghouse 3-loop PWR currently operated at rated core thermal power of 2822 MWt for each unit. The design basis for containment temperature is Main Steam Line Break (MSLB) accident at power of 2830.5 MWt, which results in peak vapor temperature of 387.6 °F. The design is such that when MSLB occurs concurrent with loss of offsite power (MSLB/LOOP), both the coolant pump on the secondary side and the fan on the air side of the FCU loose power and coast down. The pump has little inertia and coasts down in 2–3 s, while the FCU fan coasts down over much longer period. Before the pump is restored through emergency diesel generator, there is potential for boiling the coolant in the cooling coils by the high-temperature air/steam mixture entering the FCU. The time to boiling depends on the operating pressure of the coolant before the pump is restored. The prediction of the time to boiling is important because it determines whether there is potential for waterhammer or two-phase flow to occur before the pump is restored. If boiling occurs then there exists steam region in the pipe, which may cause the so called condensation induced waterhammer or column closure waterhammer. In either case, a great amount of effort has to be spent to

Technical findings and regulatory analysis for Generic Safety Issue II.E.4.3, ''Containment Integrity Check''

International Nuclear Information System (INIS)

Serkiz, A.W.

1988-04-01

This report contains the technical findings and regulatory analysis for Generic Safety Issue II.E.4.3, ''Containment Integrity Check.'' An evaluation of the containment isolation history from 1965 to 1983 reveals that (except for a small number of events) containment integrity has been maintained and that the majority of reported events have been events related to exceeding Technical Specification limits (or 0.6 of the allowable leakage level). In addition, more recent risk analyses have shown that allowable leakage rates even if increased by a factor of 10 would not significantly increase risk. Potential methods of continuous monitoring are identified and evaluated. Therefore, these technical findings and risk evaluations support closure of Generic Safety Issue II.E.4.3

Magnetic power conversion with machines containing full or porous wheel heat exchangers

Science.gov (United States)

Egolf, Peter W.; Kitanovski, Andrej; Diebold, Marc; Gonin, Cyrill; Vuarnoz, Didier

2009-04-01

A first part of the article contains a thermodynamic theory describing the temperature distribution in a Curie wheel. The occurring nonlinear ordinary differential equation has an analytical solution. If a Curie wheel is stabilized by levitation, it is named Palmy wheel. These wheels show a full structure, and because of this reason, their uptake of heat from a flame (Curie wheel) or by (solar) light absorption (Palmy wheel) only on the periphery of a cylinder is very limited. To improve the method, a modification of the principle by introducing a convective heat transport into a porous wheel is discussed. By this the power conversion rate from a heat flux to mechanical and electric power is very much increased. The second part of the article presents results of a theoretical/numerical study on the efficiencies of magnetic power conversion plants operating with porous wheels. Furthermore, these efficiencies—which are promising—are compared with those of existing power conversion plants, as e.g. geothermal binary cycle power plants.

Magnetic power conversion with machines containing full or porous wheel heat exchangers

International Nuclear Information System (INIS)

Egolf, Peter W.; Kitanovski, Andrej; Diebold, Marc; Gonin, Cyrill; Vuarnoz, Didier

2009-01-01

A first part of the article contains a thermodynamic theory describing the temperature distribution in a Curie wheel. The occurring nonlinear ordinary differential equation has an analytical solution. If a Curie wheel is stabilized by levitation, it is named Palmy wheel. These wheels show a full structure, and because of this reason, their uptake of heat from a flame (Curie wheel) or by (solar) light absorption (Palmy wheel) only on the periphery of a cylinder is very limited. To improve the method, a modification of the principle by introducing a convective heat transport into a porous wheel is discussed. By this the power conversion rate from a heat flux to mechanical and electric power is very much increased. The second part of the article presents results of a theoretical/numerical study on the efficiencies of magnetic power conversion plants operating with porous wheels. Furthermore, these efficiencies-which are promising-are compared with those of existing power conversion plants, as e.g. geothermal binary cycle power plants

Effect of seasonal variation on some physical properties and heat stability of milk subjected to ultra-high temperature and in-container sterilisation.

Science.gov (United States)

Chen, Biye; Grandison, Alistair S; Lewis, Michael J

2015-08-15

Heat stability was evaluated in bulk raw milk, collected throughout the year and subjected to ultra-high temperature (UHT) or in-container sterilisation, with and without added calcium chloride (2mM), disodium hydrogen phosphate (DSHP, 10mM) and trisodium citrate (TSC, 10mM). More sediment was observed following in-container sterilisation (0.24%) compared with UHT (0.19%). Adding CaCl2 made the milk more unstable to UHT than to in-container sterilisation, while adding DSHP and TSC made the milk more unstable during in-container sterilisation than to UHT processing, although TSC addition increased the sediment formed by UHT processing. Better heat stability was observed in autumn and winter than in spring and summer following UHT. However, following in-container sterilisation, samples with added stabilising salts showed significantly improved heat stability in autumn, whereas with added CaCl2, the best heat stability was observed in spring. No correlation was found between urea and heat stability. Copyright © 2015 Elsevier Ltd. All rights reserved.

Heat Transfer Modes and their Coefficients for a Passive Containment Cooling System of PWR using a Multi-Pod Heat Pipe

Energy Technology Data Exchange (ETDEWEB)

Nam, Gyeongho; Park, Junseok; Kim, Sangnyung [Kyung Hee Univ., Yongin (Korea, Republic of)

2013-05-15

If a reactor core is damaged due to a disaster such as happened at TEPCO's Fukushima nuclear power plant, the inevitable rise of super-heated steam that could potentially convert to hydrogen resulting from unimpeded temperature and pressure rises will threaten the integrity of the containment structure. To prevent this, safety and regulatory standards typically specify that the gas vent and external cooling systems be designed to maintain containment up to the level C limit for 24 hours and integrity for 48 hours after any damage to the core. Furthermore, it is recommended that the installation of the exhaust penetration unit have a minimum diameter of 3ft. However, installation of such cooling measures or penetration units is burdensome in terms of operational and maintenance costs not to mention the need to ensure a fleet of fire trucks to be on standby as well as the need to ensure a plentiful supply of water for cooling and a filtration system to clean the water. Therefore, the development of a reliable passive cooling system will be economically advantageous because the extra cost burdens of the external system can be omitted. The Passive Containment Cooling System (PCCS) using a multi-pod heat pipe proposed in this study satisfies these conditions.

Regulatory analyses for severe accident issues: an example

International Nuclear Information System (INIS)

Burke, R.P.; Strip, D.R.; Aldrich, D.C.

1984-09-01

This report presents the results of an effort to develop a regulatory analysis methodology and presentation format to provide information for regulatory decision-making related to severe accident issues. Insights and conclusions gained from an example analysis are presented. The example analysis draws upon information generated in several previous and current NRC research programs (the Severe Accident Risk Reduction Program (SARRP), Accident Sequence Evaluation Program (ASEP), Value-Impact Handbook, Economic Risk Analyses, and studies of Vented Containment Systems and Alternative Decay Heat Removal Systems) to perform preliminary value-impact analyses on the installation of either a vented containment system or an alternative decay heat removal system at the Peach Bottom No. 2 plant. The results presented in this report are first-cut estimates, and are presented only for illustrative purposes in the context of this document. This study should serve to focus discussion on issues relating to the type of information, the appropriate level of detail, and the presentation format which would make a regulatory analysis most useful in the decisionmaking process

Experimental analysis of heat transfer within the AP600 containment under postulated accident conditions

International Nuclear Information System (INIS)

Anderson, M.H.; Corradini, M.L.

1998-01-01

The new AP600 reactor designed by Westinghouse uses a passive safety system relying on heat removal by condensation to keep the containment within the design limits of pressure and temperature. Even though some research has been done so far in this regard, there are some uncertainties concerning the behavior of the system under postulated accident conditions. In this paper, steam condensation onto the internal surfaces of the AP600 containment walls has been investigated in two scaled vessels with similar aspect ratios to the actual AP600. The heat transfer degradation in the presence of noncondensable gas has been analyzed for different noncondensable mixtures of air and helium (hydrogen simulant). Molar fractions of noncondensables/steam ranged from (0.4-4.0) and helium concentrations in the noncondensable mixture were 0-50% by volume. In addition, the effects of the bulk temperatures, the mass fraction of noncondensable/steam, the cold wall surface temperature, the pressure, noncondensable composition, and the inclination of the condensing surface were studied. It was found that the heat transfer coefficients ranged from 50 to 800 J s -1 K -1 m -2 with the highest for high wall temperatures at high pressure and low noncondensable molar fractions. The effect of a light gas (helium) in the noncondensable mixture were found to be negligible for concentrations less than approximately 35 molar percent but could result in stratification at higher concentrations. The complete study gives a large and relatively complete data base on condensation within a scaled AP600 containment structure, providing an invaluable set of data against which to validate models. In addition, specific areas requiring further investigation are summarized. (orig.)

Shutdown decay heat removal analysis: Plant case studies and special issues: Summary report

International Nuclear Information System (INIS)

Ericson, D.M. Jr.; Cramond, W.R.; Sanders, G.A.; Hatch, S.W.

1989-04-01

Shutdown Decay Heat Removal Requirements has been designated as Unresolved Safety Issue (USI) A-45. The overall objectives of the USI A-45 program were to evaluate the safety adequacy of decay heat removal (DHR) systems in existing light water reactor nuclear power plants and to assess the value and impact (benefit-cost) of alternative measures for improving the overall reliability of the DHR function. To provide the technical data required to meet these objectives a program was developed that examined the state of DHR system reliability in a sample of existing plants. This program identified potential vulnerabilities and identified and established the feasibility of potential measures to improve the reliability of the DHR function. A value/impact (V/I) analysis of the more promising of such measures was conducted and documented. This report summarizes those studies. In addition, because of the evolving nature of V/I analyses in support of regulation, a number of supporting studies related to appropriate procedures and measures for the V/I analyses were also conducted. These studies are also summarized herein. This report only summarizes findings of technical studies performed by Sandia National Laboratories as part of the program to resolve this issue. 46 refs., 7 figs., 124 tabs

Atmospheric stability inside containments with a heated layer of liquid on the floor

Energy Technology Data Exchange (ETDEWEB)

Vate, J.F. van de [Netherlands Energy Research Foundation, Petten (Netherlands)

1977-01-01

The study of atmospheric stability inside containments with a heated layer of liquid comprised derivation of the boundary condition for stable atmospheric stratifications and the experimental validation of the boundary condition for stable atmospheric stratification. This report includes description of the model for stirred aerosol deposition and the calculation results for maximum aerodynamic diameter of a confined aerosol remaining just well-stirred.

Design, materials and R and D issues of innovative thermal contact joints for high heat flux applications

International Nuclear Information System (INIS)

Federici, G.; Haines, J.; Tillack, M.S.; Ulrickson, M.

1995-01-01

Plasma facing components in fusion machines are designed with a layer of sacrificial armour material facing the plasma and a high-conductivity material in contact with the coolant. One of the most critical issues associated with making the proposed design concept work, from a power handling point of view, is achieving the necessary contact conductance between the armour and the heat sink.This paper presents a novel idea for the interface joint between the sacrificial armour and the actively cooled permanent heat sink. It consists of a thermal bond layer of a binary or more complex alloy, treated in the semi-solid region in such a way as to lead to a fine dispersion of a globular solid phase into a liquid matrix (rheocast process). The alloy in this ''mushy state'' exhibits a time-dependent, shear rate-dependent viscosity, which is maintained reversibly when the material is solidified and heated again in the semi-solid state. The function of the thermal bond layer is to facilitate heat transfer between the replaceable armour and the permanent heat sink without building up excessive thermal stresses, as in conventional brazed joints, and allow an easy replacement whenever needed without disturbing the coolant system. No contact pressure is required in this case to provide the desired heat transfer conductance, and the reversible thixotropic properties of the rheocast material should guarantee the stability of the layer in the semi-solid conditions.Key design, material and testing issues are identified and discussed in this paper with emphasis on specific needs for future research and development work. Examples of suitable material options which are being considered are reported together with some initial heat transfer analysis results. (orig.)

Heat transfer properties of organic coolants containing high boiling residues

International Nuclear Information System (INIS)

Debbage, A.G.; Driver, M.; Waller, P.R.

1964-01-01

Heat transfer measurements were made in forced convection with Santowax R, mixtures of Santowax R and pyrolytic high boiling residue, mixtures of Santowax R and CMRE Radiolytic high boiling residue, and OMRE coolant, in the range of Reynolds number 10 4 to 10 5 . The data was correlated with the equation Nu = 0.015 Re b 0.85 Pr b 0.4 with an r.m.s. error of ± 8.5%. The total maximum error arising from the experimental method and inherent errors in the physical property data has been estimated to be less than ± 8.5%. From the correlation and physical property data, the decrease in heat transfer coefficient with increasing high boiling residue concentration has been determined. It has been shown that subcooled boiling in organic coolants containing high boiling residues is a complex phenomenon and the advantages to be gained by operating a reactor in this region may be marginal. Gas bearing pumps used initially in these experiments were found to be unsuitable; a re-designed ball bearing system lubricated with a terphenyl mixture was found to operate successfully. (author)

Microstructure and heat resistance of Mg-Al-Zn alloys containing metastable phase

International Nuclear Information System (INIS)

Kim, Jeong-Min; Park, Bong-Koo; Jun, Joong-Hwan; Shin, Keesam; Kim, Ki-Tae; Jung, Woon-Jae

2007-01-01

In this research microstructural studies have been made on cast specimens of AZ91 base alloys containing various amounts of Zn. As the amount of Zn addition increased up to 2%, any new Zn-containing phase did not appear while the Zn content in Mg 17 Al 12 phase continuously increased. A quasi-crystalline phase started to form at Mg 17 Al 12 phase when the added Zn content was about 3 wt.%. The tensile strength and elongation of the alloys at 175 deg. C were observed to increase significantly with increasing Zn content. The quasi-crystalline phase was found to be stable up to 300 deg. C, based on scanning electron microscopy examinations of the specimens heated at different temperatures for 24 h

Results from the DCH-1 [Direct Containment Heating] experiment

International Nuclear Information System (INIS)

Tarbell, W.W.; Brockmann, J.E.; Pilch, M.; Ross, J.E.; Oliver, M.S.; Lucero, D.A.; Kerley, T.E.; Arellano, F.E.; Gomez, R.D.

1987-05-01

The DCH-1 (Direct Containment Heating) test was the first experiment performed in the Surtsey Direct Heating Test Facility. The test involved 20 kg of molten core debris simulant ejected into a 1:10 scale model of the Zion reactor cavity. The melt was produced by a metallothermic reaction of iron oxide and aluminum powders to yield molten iron and alumina. The cavity model was placed so that the emerging debris propagated directly upwards along the vertical centerline of the chamber. Results from the experiment showed that the molten material was ejected from the caviity as a cloud of particles and aerosol. The dispersed debris caused a rapid pressurization of the 103-m 3 chamber atmosphere. Peak pressure from the six transducers ranged from 0.09 to 0.13 MPa (13.4 to 19.4 psig) above the initial value in the chamber. Posttest debris collection yielded 11.6 kg of material outside the cavity, of which approximately 1.6 kg was attributed to the uptake of oxygen by the iron particles. Mechanical sieving of the recovered debris showed a lognormal size distribution with a mass mean size of 0.55 mm. Aerosol measurements indicated a subsantial portion (2 to 16%) of the ejected mass was in the size range less than 10 m aerodynamic equivalent diameter

Environmental issues and competitiveness of district heating systems

International Nuclear Information System (INIS)

Kypreos, S.

1991-01-01

The advantages of district heating systems are evaluated in competition to individual heating for the Swiss markets. The preservation of the environmental quality on the national (clean air concept) and global scale (Toronto recommendation) is formulated as constraint of the energy system. The implications of these constraints for the economic competition of district heating is evaluated. The study estimates the evolution of energy demand in the heating markets and shortly describes the technical possibilities in satisfying demand by a set of conventional heating systems, systems using renewable energy sources, energy conservation measures and district heating systems based on conventional or nuclear energy sources. The main conclusion is that small capacity nuclear district heating systems, if acceptable, could enhance the flexibility of the Swiss energy system in respect to CO 2 control. (author) 3 figs., 4 tabs., 9 refs

Standard Practice for Laboratory Screening of Metallic Containment Materials for Use With Liquids in Solar Heating and Cooling Systems

CERN Document Server

American Society for Testing and Materials. Philadelphia

1980-01-01

1.1 This practice covers several laboratory test procedures for evaluating corrosion performance of metallic containment materials under conditions similar to those that may occur in solar heating and cooling systems. All test results relate to the performance of the metallic containment material only as a part of a metal/fluid pair. Performance in these laboratory test procedures, taken by itself, does not necessarily constitute an adequate basis for acceptance or rejection of a particular metal/fluid pair in solar heating and cooling systems, either in general or in a particular design. This practice is not intended to preclude the use of other screening tests, particularly when those tests are designed to more closely simulate field service conditions. 1.2 This practice describes apparatus and procedures for several tests, any one or more of which may be used to evaluate the deterioration of the metallic containment material in a metal/fluid pair. The procedures are designed to permit simulation, heating...

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.)

Storage vessel for radiation contaminated container

International Nuclear Information System (INIS)

Sakatani, Tadatsugu.

1996-01-01

In a storage vessel of the present invention, a plurality of radiation contaminated material containing bodies are vertically stacked in a cell chamber. Then, the storage vessel comprises a containing tube for containing a plurality of the containing bodies, cooling coils wound around the containing tube, a cooling medium circulating system connected to the cooling coils and circulating cooling medium, and a heat exchanger interposed to the cooling medium circulating system for removing heat of the cooling medium. Heat of the radioactive material containing bodies is transferred to cooling air and cooling coils by way of the container tube, thereby cooling the containing bodies. By the operation of circulating pumps in a cooling medium circulation system, the cooling medium circulates through a circulation channel comprising a cooling medium transfer pipes, cooling medium branching tubes, cooling coils and the heat exchanger, then heat of the cooling medium is transferred to a heat utilizing system by way of the heat exchanger to attain effective utilization of the heat. In this case, heat can be taken out stably even when the storage amount fluctuates and heat releasing amount is reduced, and improvement of heat transfer promotes the cooling of the containing bodies, which enables minimization of the size of the storage vessel. (T.M.)

Heat transfer from internally heated hemispherical pools

International Nuclear Information System (INIS)

Gabor, J.D.; Ellsion, P.G.; Cassulo, J.C.

1980-01-01

Experiments were conducted on heat transfer from internally heated ZnSO 4 -H 2 O pools to the walls of hemispherical containers. This experimental technique provides data for a heat transfer system that has to date been only theoretically treated. Three different sizes of copper hemispherical containers were used: 240, 280, 320 mm in diameter. The pool container served both as a heat transfer surface and as an electrode. The opposing electrode was a copper disk, 50 mm in diameter located at the top of the pool in the center. The top surface of the pool was open to the atmosphere

Fundamental study on the melting process of crushed ice in a heat storage container; Chikunetsu sonai ni takuwaeta saihyo no yukai ni kansuru kisoteki kenkyu

Energy Technology Data Exchange (ETDEWEB)

Yanadori, M; Kobori, H [Hitachi, Ltd., Tokyo (Japan); Tsubota, Y [Tokyo Electric Power Co. Inc., Tokyo (Japan)

1998-03-25

This report deals with heat transfer in the melting process of crushed ice filling in a ice/water heat storage container. Volumetric heat transfer rate and melting end-time are measured when rectangular-type, small-stone-type and particle-type ice in the container are melted by circulation hot water. Melting end-time of small-stone-type ice is the shortest and that of particle-type ice is the latest. Volumetric heat transfer rate of small-stone-type ice and rectangular-type ice is larger than that of particle-type ice. The flow rate of circulation hot water throwing in container through a inlet pipe influences remarkably on heat transfer rate. 4 refs., 10 figs.

Heat Transfer and Latent Heat Storage in Inorganic Molten Salts for Concentrating Solar Power Plants

Energy Technology Data Exchange (ETDEWEB)

Mathur, Anoop [Terrafore Inc.

2013-08-14

A key technological issue facing the success of future Concentrating Solar Thermal Power (CSP) plants is creating an economical Thermal Energy Storage (TES) system. Current TES systems use either sensible heat in fluids such as oil, or molten salts, or use thermal stratification in a dual-media consisting of a solid and a heat-transfer fluid. However, utilizing the heat of fusion in inorganic molten salt mixtures in addition to sensible heat , as in a Phase change material (PCM)-based TES, can significantly increase the energy density of storage requiring less salt and smaller containers. A major issue that is preventing the commercial use of PCM-based TES is that it is difficult to discharge the latent heat stored in the PCM melt. This is because when heat is extracted, the melt solidifies onto the heat exchanger surface decreasing the heat transfer. Even a few millimeters of thickness of solid material on heat transfer surface results in a large drop in heat transfer due to the low thermal conductivity of solid PCM. Thus, to maintain the desired heat rate, the heat exchange area must be large which increases cost. This project demonstrated that the heat transfer coefficient can be increase ten-fold by using forced convection by pumping a hyper-eutectic salt mixture over specially coated heat exchanger tubes. However,only 15% of the latent heat is used against a goal of 40% resulting in a projected cost savings of only 17% against a goal of 30%. Based on the failure mode effect analysis and experience with pumping salt at near freezing point significant care must be used during operation which can increase the operating costs. Therefore, we conclude the savings are marginal to justify using this concept for PCM-TES over a two-tank TES. The report documents the specialty coatings, the composition and morphology of hypereutectic salt mixtures and the results from the experiment conducted with the active heat exchanger along with the lessons learnt during

Study on the application of thickened welds without post weld heat treatment for containment vessels

International Nuclear Information System (INIS)

Takeuchi, T.; Fukaya, T.; Sato, M.; Takano, G.

1978-01-01

As material for containment vessels, SGV49 steel plates are mainly used. However, those used for this purpose are limited in thickness to smaller than 38 mm. This is because the present standard requires welds thicker than 38 mm to be subjected to post weld heat treatment but operation on the site is practically difficult. In the case of 3-loop containment vessels of pressurized water type reactors, use of 38 mm SGV49 brings an increase in their height and this is disadvantageous from a seismic viewpoint. Therefore, use of 45 mm-thick steel material has become necessary in order to increase design internal pressure and reduce the height of the vessels. To investigate the propriety of the use of 45 mm-thick SGV49 for this purpose without post weld heat treatment we investigated the basic performances of base metal and welded joints. We also conducted large-scale embrittlement fracture tests (CT test, deep notch test, wide plate tensile test and ESSO test) in order to examine whether welds not subjected to post weld heat treatment are safe against embrittlement fracture under the operating conditions of the vessels. The results proved that the welds of SGV49 steel plates are safe enough under the operating conditions. (author)

Bringing a probiotic-containing functional food to the market: microbiological, product, regulatory and labeling issues.

Science.gov (United States)

Sanders, M E; Huis in't Veld, J

1999-01-01

Properly formulated probiotic-containing foods offer consumers a low risk, low cost dietary component that has the potential to promote health in a variety of ways. Several such products are available commercially, although markets in Japan and Europe are more developed than in the USA. Once healthful attributes of a probiotic product have been identified, there remain microbiological, product, regulatory and labeling issues to be addressed prior to marketing. Microbiological and product issues include safety, effective scale-up for manufacturing, definition of probiotic activity, probiotic stability in the product over the course of product manufacture, shelf-life and consumption, definition of effective dose and target population(s), and development of quality assurance approaches. Examples of probiotic-containing foods are given. Regulatory and labeling issues are complicated because they differ for each country, but are likewise critical because they provide the means for communication of the product benefits to the consumer. The regulatory climate worldwide appears to be one of caution about overstating the benefits of such products but at the same time not preventing corporate commitment to marketing.

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.)

Experiments to investigate direct containment heating phenomena with scaled models of the Surry Nuclear Power Plant

International Nuclear Information System (INIS)

Blanchat, T.K.; Allen, M.D.; Pilch, M.M.

1994-01-01

The Containment Technology Test Facility (CTTF) and the Surtsey Test Facility at Sandia National Laboratories (SNL) are used to perform scaled experiments for the Nuclear Regulatory Commission (NRC) that simulate High Pressure Melt Ejection (HPME) accidents in a nuclear power plant (NPP). These experiments are designed to investigate the effects of direct containment heating (DCH) phenomena on the containment load. High-temperature, chemically reactive melt is ejected by high-pressure steam into a scale model of a reactor cavity. Debris is entrained by the steam blowdown into a containment model where specific phenomena, such as the effect of subcompartment structures, prototypic atmospheres, and hydrogen generation and combustion, can be studied

GOTHIC-IST 6.1b code validation exercises relating to heat removal by dousing and air coolers in CANDU containment

International Nuclear Information System (INIS)

Ramachandran, S.; Krause, M.; Nguyen, T.

2003-01-01

This paper presents validation results relating to the use of the GOTHIC containment analysis code for CANDU safety analysis. The validation results indicate that GOTHIC predicts heat removal by dousing and air cooler heat transfer with reasonable accuracy. (author)

Evaluation of conceptual Heat Exchanger Design for passive containment cooling system of SMART

Energy Technology Data Exchange (ETDEWEB)

Kim, Min-Ki; Hong, Soon Joon [FNC Tech., Yongin (Korea, Republic of); Kim, Young In; Kim, Seok [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2016-10-15

PCCS(Passive containment cooling system) is the passive safety system which ultimately removes the reactor decay heat. Cooling performance of the air-cooled type and water-circulation cooling type of PCCS were analyzed using CAP version 2.21. The analysis results show the water-circulation cooling PCCS is more effective in lowering the peak pressure and temperature in the containment building. However, the air-cooled PCCS is more effective to the long-term cooling. From this study, the efficiency evaluation results for the two PCCS designs are obtained. These results may be applied in the PCCS design improvement. Moreover, these results will be used as a reference for the later PCCS design and analysis.

Heat transfer simulation and retort program adjustment for thermal processing of wheat based Haleem in semi-rigid aluminum containers.

Science.gov (United States)

Vatankhah, Hamed; Zamindar, Nafiseh; Shahedi Baghekhandan, Mohammad

2015-10-01

A mixed computational strategy was used to simulate and optimize the thermal processing of Haleem, an ancient eastern food, in semi-rigid aluminum containers. Average temperature values of the experiments showed no significant difference (α = 0.05) in contrast to the predicted temperatures at the same positions. According to the model, the slowest heating zone was located in geometrical center of the container. The container geometrical center F0 was estimated to be 23.8 min. A 19 min processing time interval decrease in holding time of the treatment was estimated to optimize the heating operation since the preferred F0 of some starch or meat based fluid foods is about 4.8-7.5 min.

Specific heat jump at T/sub c/ of proximity effect sandwiches containing nonmagnetic localized states

International Nuclear Information System (INIS)

Maneeratankul, S.; Tang, I.M.

1987-01-01

The decrease in the transition temperature and the jump in the specific heat at T/sub c/ of proximity effect sandwiches containing nonmagnetic Anderson impurities in the normal layer are studied. The effects of the resonant scattering by the impurities are treated in the same manner as that used by Kaiser in his study of the effects of resonant scattering on the properties of bulk superconductors. Numerical calculations of the decrease in T/sub c/ and the jump in the specific heat at T/sub c/ as a function of the thickness of the normal layer are presented

Flammable gas issues in double-contained receiver tanks. Revision 1

International Nuclear Information System (INIS)

Peurrung, L.M.; Mahoney, L.A.; Stewart, C.W.; Gauglitz, P.A.; Pederson, L.R.; Bryan, S.A.; Shepard, C.L.

1998-06-01

Four double-contained receiver tanks (DCRTs) at Hanford will be used to store salt-well pumped liquids from tanks on the Flammable Gas Watch List. This document was created to serve as a technical basis or reference document for flammable gas issues in DCRTs. The document identifies, describes, evaluates, and attempts to quantify potential gas carryover and release mechanisms. It estimates several key parameters needed for these calculations, such as initial aqueous concentrations and ventilation rate, and evaluates the uncertainty in those estimates. It justifies the use of the Schumpe model for estimating vapor-liquid equilibrium constants. It identifies several potential waste compatibility issues (such as mixing and pH or temperature changes) that could lead to gas release and provides a basis for calculating their effects. It evaluates the potential for gas retention in precipitated solids within a DCRT and whether retention could lead to a buoyant displacement instability (rollover) event. It discusses rates of radiolytic, thermal, and corrosive hydrogen generation within the DCRT. It also describes in detail the accepted method of calculating the lower flammability limit (LFL) for mixtures of flammable gases

Flammable gas issues in double-contained receiver tanks. Revision 1

Energy Technology Data Exchange (ETDEWEB)

Peurrung, L.M.; Mahoney, L.A.; Stewart, C.W.; Gauglitz, P.A.; Pederson, L.R.; Bryan, S.A.; Shepard, C.L.

1998-06-01

Four double-contained receiver tanks (DCRTs) at Hanford will be used to store salt-well pumped liquids from tanks on the Flammable Gas Watch List. This document was created to serve as a technical basis or reference document for flammable gas issues in DCRTs. The document identifies, describes, evaluates, and attempts to quantify potential gas carryover and release mechanisms. It estimates several key parameters needed for these calculations, such as initial aqueous concentrations and ventilation rate, and evaluates the uncertainty in those estimates. It justifies the use of the Schumpe model for estimating vapor-liquid equilibrium constants. It identifies several potential waste compatibility issues (such as mixing and pH or temperature changes) that could lead to gas release and provides a basis for calculating their effects. It evaluates the potential for gas retention in precipitated solids within a DCRT and whether retention could lead to a buoyant displacement instability (rollover) event. It discusses rates of radiolytic, thermal, and corrosive hydrogen generation within the DCRT. It also describes in detail the accepted method of calculating the lower flammability limit (LFL) for mixtures of flammable gases.

Container for spent fuel assembly

International Nuclear Information System (INIS)

Sawai, Takeshi.

1996-01-01

The container of the present invention comprises a container main body having a body portion which can contain spent fuel assemblies and a lid, and heat pipes having an evaporation portion disposed along the outer surface of the spent fuel assemblies to be contained and a condensation portion exposed to the outside of the container main body. Further, the heat pipe is formed spirally at the evaporation portions so as to surround the outer circumference of the spent fuel assemblies, branched into a plurality of portions at the condensation portion, each of the branched portion of the condensation portion being exposed to the outside of the container main body, and is tightly in contact with the periphery of the slit portions disposed to the container main body. Then, since released after heat is transferred to the outside of the container main body from the evaporation portion of the heat pipe along the outer surface of the spent fuel assemblies by way of the condensation portion of the heat pipes exposed to the outside of the container main body, the efficiency of the heat transfer is extremely improved to enhance the effect of removing heat of spent fuel assemblies. Further, cooling effect is enhanced by the spiral form of the evaporation portion and the branched condensation portion. (N.H.)

Thermodynamic analysis on an instantaneous water heating system of shower wastewater source heat pump

Directory of Open Access Journals (Sweden)

Yuguo Wu

2018-09-01

Full Text Available Water reuse and desalination systems are energy intensive processes, and their increasing use is leading energy consumption within water systems to be an increasingly important issue. Shower wastewater contains large amounts of heat, so there is an opportunity to recover energy from shower water to offset energy consumption elsewhere in water systems. This paper found ways to increase the output of hot water and lower the energy consumption by establishing a thermodynamic model of an instantaneous wastewater source heat pump. The system proved to be very effective, the heating COP (coefficient of performance can reach 3.3 even in the winter. Under the conditions of limited heat transfer area, reducing the suction pressure of a compressor is a more feasible way to increase the hot water output to meet the needs of users rather than increasing the discharge pressure. Besides, increasing the heat transfer area of the evaporator is a more effective option. When the heat transfer area of evaporator varies from 0.5 to 1.0 square meters, a notable change is that the heating COP increases from 3.283 to 3.936. The heating COP in a system with a recuperator can reach 5.672, almost double that compared to the original systems.

Numerical analysis of heat transfer of canned liquid foods containing fibers or particles during sterilization

Energy Technology Data Exchange (ETDEWEB)

Wang, Q.Z.; Sakai, N.; Hanzawa, T. [Tokyo Univ. of Fisheries, Tokyo (Japan). Dept. of Food Science and Tech.

2000-10-01

The velocity profile, temperature distribution, and the slowest heating point of a canned liquid food containing fibers or particles were calculated numerically by using fundamental equations that take account of the effect of free convection in the can at an unsteady state under the assumption of imaginary fluid with apparent physical properties. To check these calculated results, the temperature distribution in the can was measured experimentally under the same operating conditions as those of the theoretical analysis. The calculated results agree closely with the experimental ones. Adaptable ranges of present numerical analysis and the positional characteristics of the slowest heating point are shown. (author)

New hybrid nanofluid containing encapsulated paraffin wax and sand nanoparticles in propylene glycol-water mixture: Potential heat transfer fluid for energy management

International Nuclear Information System (INIS)

Manikandan, S.; Rajan, K.S.

2017-01-01

Highlights: • Hybrid nanofluid containing sand nanoparticles & encapsulated paraffin wax prepared. • Specific heat of hybrid nanofluid 9% greater than that of PG-water mixture. • Specific heat & thermal conductivity enhanced at optimum paraffin wax concentration. • Hybrid nanofluid with 1 wt.% paraffin wax & 1 vol% sand nanoparticles best suited. - Abstract: The reduction in specific heat commonly encountered due to the addition of nanoparticles to a heat transfer fluid such as propylene glycol-water mixture, can be overcome by co-dispersing surfactant-encapsulated paraffin wax, leading to formation of a hybrid nanofluid. Experimental investigations have been carried out on the preparation and evaluation of thermophysical properties of a hybrid nanofluid containing pluronic P-123 encapsulated paraffin wax (70–120 nm diameter, 1–5 wt.%) and sand nanoparticles (1 vol%) in propylene glycol-water mixture. The comparison of results of differential scanning calorimetry of pure paraffin wax and encapsulated paraffin wax revealed encapsulation efficiency of 84.4%. The specific heat of hybrid nanofluids monotonously increased with paraffin wax concentration, with 9.1% enhancement in specific heat for hybrid nanofluid containing 5 wt.% paraffin wax, in comparison to propylene glycol-water mixture. There exists an optimum paraffin wax concentration (1 wt.%) for the hybrid nanofluid at which the combination of various thermophysical properties such as specific heat, thermal conductivity and viscosity are favorable for use as heat transfer fluid. Such a hybrid nanofluid can be used as a substitute for propylene glycol-water mixture in solar thermal systems.

Passive cooling containment study

International Nuclear Information System (INIS)

Shin, J.J.; Iotti, R.C.; Wright, R.F.

1993-01-01

Pressure and temperature transients of nuclear reactor containment following postulated loss of coolant accident with a coincident station blackout due to total loss of all alternating current power are studied analytically and experimentally for the full scale NPR (New Production Reactor). All the reactor and containment cooling under this condition would rely on the passive cooling system which removes reactor decay heat and provides emergency core and containment cooling. Containment passive cooling for this study takes place in the annulus between containment steel shell and concrete shield building by natural convection air flow and thermal radiation. Various heat transfer coefficients inside annular air space were investigated by running the modified CONTEMPT code CONTEMPT-NPR. In order to verify proper heat transfer coefficient, temperature, heat flux, and velocity profiles were measured inside annular air space of the test facility which is a 24 foot (7.3m) high, steam heated inner cylinder of three foot (.91m) diameter and five and half foot (1.7m) diameter outer cylinder. Comparison of CONTEMPT-NPR and WGOTHIC was done for reduced scale NPR

Storage chamber for container of radiation-contaminated material

International Nuclear Information System (INIS)

Takakura, Masahide.

1996-01-01

The present invention concerns a storage chamber for containing radiation-contaminated materials in containing tubes and having cooling fluids circulated at the outer side of the containing tubes. The storage chamber comprises a gas supply means connected to the inside of the container tube for supplying a highly heat-conductive gas and a gas exhaustion means for discharging the gas present in the container tube. When containing vessels for radiation-contaminated materials are contained in the container tube, the gases present inside of the container tube is exhausted by means of the gas exhaustion means, and highly heat conductive gases are filled from the gas supply means to the space between the container tube and the containing vessels for the radiation-contaminated materials. When the temperature of the highly heat conductive gas is elevated due to the heat generation of the radiation-contaminated materials, the container tube is heated, and then cooled by the cooling fluid at the outer side of the container tube. In this case, the heat of the radiation-contaminated material-containing vessels is removed by the heat conduction by the highly heat conductive gas to reduce temperature gradient between the containing vessels and the containing tube. This can enhance the cooling effect. (T.M.)

Containing method for spent fuel and spent fuel containing vessel

International Nuclear Information System (INIS)

Maekawa, Hiromichi; Hanada, Yoshine.

1996-01-01

Upon containing spent fuels, a metal vessel main body and a support spacer having fuel containing holes are provided. The support spacer is disposed in the inside of the metal vessel main body, and spent fuel assemblies are loaded in the fuel containing holes. Then, a lid is welded at the opening of the metal vessel main body to provide a sealing state. In this state, heat released from the spent fuel assemblies is transferred to the wall of the metal vessel main body via the support spacer. Since the support spacer has a greater heat conductivity than gases, heat of the spent fuel assemblies tends to be released to the outside, thereby capable of removing heat of the spent fuel assemblies effectively. In addition, since the surfaces of the spent fuel assemblies are in contact with the inner surface of the fuel containing holes of the support spacer, impact-resistance and earthquake-resistance are ensured, and radiation from the spent fuel assemblies is decayed by passing through the layer of the support spacer. (T.M.)

Effect of prolonged heating on the asphalt-aggregate bond strength of HMA containing liquid anti-strip additives

Science.gov (United States)

2008-09-01

In this study, an attempt was made to determine the effect of prolonged heating on the bond strength between : aggregate and asphalt that contained anti-strip additives (LOF 6500 and Morelife 2200). On account of the : substantial decrease of anti-st...

GOTHIC MODEL OF BWR SECONDARY CONTAINMENT DRAWDOWN ANALYSES

International Nuclear Information System (INIS)

Hansen, P.N.

2004-01-01

This article introduces a GOTHIC version 7.1 model of the Secondary Containment Reactor Building Post LOCA drawdown analysis for a BWR. GOTHIC is an EPRI sponsored thermal hydraulic code. This analysis is required by the Utility to demonstrate an ability to restore and maintain the Secondary Containment Reactor Building negative pressure condition. The technical and regulatory issues associated with this modeling are presented. The analysis includes the affect of wind, elevation and thermal impacts on pressure conditions. The model includes a multiple volume representation which includes the spent fuel pool. In addition, heat sources and sinks are modeled as one dimensional heat conductors. The leakage into the building is modeled to include both laminar as well as turbulent behavior as established by actual plant test data. The GOTHIC code provides components to model heat exchangers used to provide fuel pool cooling as well as area cooling via air coolers. The results of the evaluation are used to demonstrate the time that the Reactor Building is at a pressure that exceeds external conditions. This time period is established with the GOTHIC model based on the worst case pressure conditions on the building. For this time period the Utility must assume the primary containment leakage goes directly to the environment. Once the building pressure is restored below outside conditions the release to the environment can be credited as a filtered release

Technical specification improvements to containment heat removal and emergency core cooling systems: Final report

International Nuclear Information System (INIS)

Sullivan, W.P.; Ha, C.; Pentzien, D.C.; Visweswaran, S.

1988-07-01

This report presents the results of an analysis for technical specification improvements to the emergency core cooling systems (ECCS) and containment heat removal systems (EPRI Research Project 2142-3). The objective of this project is to further develop a reliability- and risk-based methodology to provide improvements by considering groups of surveillance test intervals and allowed out-of-service times jointly. This was done for the technical specifications for the ECCS, containment heat removal equipment, and supporting systems of a boiling water reactor plant. The project (1) developed a methodology for optimizing groups of surveillance test intervals and allowed out-of-service times jointly, (2) applied the methodology in a case study of a specific operating plant, Hatch-2, and (3) evaluated benefits of the application. The results of the case study demonstrate that beneficial technical specification improvements can be realized with application of the methodology. By tightening a small group of sensitive surveillance test intervals (STIs) and allowed out-of-service times (AOTs), a larger group of less sensitive STIs and AOTs can be extended resulting in an overall plant operating cost improvement without reducing the plant safety. The reliability- and risk-based methodology and results from this project can be effectively applied for technical specification improvements at other operating plants

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

Compact interior heat exchangers for CO{sub 2} mobile heat pumping systems

Energy Technology Data Exchange (ETDEWEB)

Hafner, Armin

2003-07-01

The natural refrigerant carbon dioxide (CO{sub 2}) offers new possibilities for design of flexible, efficient and environmentally safe mobile heat pumping systems. As high-efficient car engines with less waste heat are developed, extra heating of the passenger compartment is needed in the cold season. A reversible transcritical CO{sub 2} system with gliding temperature heat rejection can give high air delivery temperature which results in rapid heating of the passenger compartment and rapid defogging or defrosting of windows. When operated in cooling mode, the efficiency of transcritical CO{sub 2} systems is higher compared to common (HFC) air conditioning systems, at most dominant operating conditions. Several issues were identified for the design of compact interior heat exchangers for automotive reversible CO{sub 2} heat pumping systems. Among theses issues are: (1) Refrigerant flow distribution, (2) Heat exchanger fluid flow circuiting, (3) Air temperature uniformity downstream of the heat exchanger, (4) Minimization of temperature approach, (5) Windshield flash fogging due to retained water inside the heat exchanger, (6) Internal beat conduction in heating mode operation, and (7) Refrigerant side pressure drop In order to provide a basis for understanding these issues, the author developed a calculation model and set up a test facility and investigated different prototype heat exchangers experimentally.

Application of a two-cell adiabatic model for direct containment heating to the ABB C-E system 80+ ALWR

International Nuclear Information System (INIS)

Schneider, R.E.; Sherry, R.R.

1993-01-01

During certain severe reactor accidents, such as those initiated by a station blackout or small-break loss of coolant accident (LOCA) degradation of the reactor core can take place while the reactor coolant system remains pressurized. If unmitigated, core materials will melt and relocate to the lower regions of the reactor pressure vessel and ultimately melt through the reactor pressure vessel (RPV) lower head. Once the RPV is breached, core debris will be ejected from the RPV and entrained from the reactor cavity by the high velocity gases blowing down from the reactor vessel. During the entrainment process, metallic constituents of the ejected material, principally zirconium and steel, exothermically react with oxygen and steam to generate chemical energy and (in the case of reactions with steam) hydrogen. Concomitant with the high pressure melt ejection (HPME) process, there is the potential for hydrogen combustion and vaporization of available water. The sensible heat loss to the containment atmosphere and the associated processes are typically referred to as direct containment heating (DCH). If large quantities of energy from the corium and corium-steam reactions are transferred directly to the containment atmosphere, the containment may pressurize to a point where failure is possible. Since the containment threat is coincident with vessel breach, relatively high containment radiation releases would be expected from this type of containment failure

Implementation of Exhaust Gas Recirculation for Double Stage Waste Heat Recovery System on Large Container Vessel

DEFF Research Database (Denmark)

Andreasen, Morten; Marissal, Matthieu; Sørensen, Kim

2014-01-01

Concerned to push ships to have a lower impact on the environment, the International Maritime Organization are implementing stricter regulation of NOx and SOx emissions, called Tier III, within emission control areas (ECAs). Waste Heat Recovery Systems (WHRS) on container ships consist...... of recovering some of the waste heat from the exhaust gas. This heat is converted into electrical energy used on-board instead of using auxiliary engines. Exhaust Gas Recirculation (EGR) systems, are recirculating a part of the exhaust gas through the engine combustion chamber to reduce emissions. WHRS combined...... with EGR is a potential way to improve system efficiency while reducing emissions. This paper investigates the feasibility of combining the two systems. EGR dilutes the fuel, lowering the combustion temperature and thereby the formation of NOx, to reach Tier III limitation. A double stage WHRS is set up...

Heat Shock Protein-Inducing Property of Diarylheptanoid Containing Chalcone Moiety from Alpinia katsumadai

Directory of Open Access Journals (Sweden)

Joo-Won Nam

2017-10-01

Full Text Available A new diarylheptanoid containing a chalcone moiety, katsumain H (1, was isolated from the seeds of Alpinia katsumadai. The structure was elucidated using a combination of 1D/2D NMR spectroscopy and mass spectrometry data analysis. The absolute configurations of C-3, C-5, and C-7 in 1 were assigned based on its optical rotation and after comparing its NMR chemical shifts with those of its diastereoisomers, katsumain E and katsumain F, which were previously isolated from this plant and characterized. In this study, the stimulatory effects of compounds 1 and 2 were evaluated on heat shock factor 1 (HSF1, heat shock protein 27 (HSP27, and HSP70. Compounds 1 and 2 increased the expression of HSF1 (1.056- and 1.200-fold, respectively, HSP27 (1.312- and 1.242-fold, respectively, and HSP70 (1.234- and 1.271-fold, respectively, without increased cytotoxicity.

In situ vitrification of buried waste: Containment issues and suppression systems

International Nuclear Information System (INIS)

Luey, J.; Powell, T.D.

1992-03-01

Pacific Northwest Laboratory (PNL) and Idaho National Engineering Laboratory (INEL) are developing a remedial action technology for buried waste through the adaptation of the in situ vitrification (ISV) process. The ISV process is a thermal treatment process originally developed for the US Department of Energy (DOE) to stabilize soils contaminated with transuranic waste. ISV tests with buried waste forms have demonstrated that the processing of buried waste is more dynamic than the processing of soils. This paper will focus on the issue of containment of the gases released during the processing of buried waste and on engineered suppression systems to alleviate transient events associated with dynamic off-gassing from the ISV melt

In situ vitrification of buried waste: Containment issues and suppression systems

International Nuclear Information System (INIS)

Luey, J.; Powell, T.D.

1992-01-01

Pacific Northwest Laboratory (PNL) and Idaho National Engineering Laboratory (INEL) are developing a remedial action technology for buried waste through the adaptation of the in situ vitrification (ISV) process. The ISV process is a thermal treatment process originally developed for the U.S. Department of Energy (DOE) to stabilize soils contaminated with transuranic waste. ISV tests with buried waste forms have demonstrated that the processing of buried waste is more dynamic than the processing of soils. This paper will focus on the issue of containment of the gases released during the processing of buried waste and on engineered suppression systems to alleviate transient events associated with dynamic off-gassing from the ISV melt. (author)

Heat transfer characteristics of UF6 in a container heated from outer surface. Pt. 1. Thermal hydraulic analysis method taking account of phase change and volume expansion

International Nuclear Information System (INIS)

Wataru, Masumi; Gomi, Yoshio; Yamakawa, Hidetsugu; Tsumune, Daisuke

1995-01-01

Natural UF6 is transported in a steel container from foreign countries to the enrichment plant in Japan. If the container meets fire accident, it is heated by fire (800degC) and rupture of the container may occur. For the safety point of view, it is necessary to know whether rupture occurs or not. Because UF6 has a radiological and chemical hazards, it is difficult to perform a demonstration test with UF6. So thermal calculation method has to be developed. The rupture is caused by UF6 gaseous pressure or volume expansion of liquid UF6. To know time history of internal pressure and temperature distribution in the container, it is important to evaluate thermal phenomena of UF6. When UF6 is heated, it changes from solid to liquid or gas at low temperature (64degC) and then its volume expands little by little. In this study, thermal calculation method has been developed taking phase change and thermal expansion of UF6 into account. In the calculation, a two-dimensional model is adopted and natural convection of liquid UF6 is analyzed. As a result of this study, numerical solutions have been obtained taking phase change and volume expansion into account. (author)

Deposition of aerosols formed by HCDA due to decay heat transport in inner containment atmospheres

International Nuclear Information System (INIS)

Vate, J.F. van de

1976-01-01

Coupling of decay heat transfer by aerosol-laden inner containment atmospheres with aerosol deposition from such atmospheres leads to useful and simple models for calculation of the time dependence of the aerosol mass concentration. Special attention is given to thermophoretic deposition (dry case) and condensation followed by gravitational deposition (wet case). Attractive features of the models are: 1) coagulation can be omitted and therefore complicated and doubtful calculations on coagulation are avoided, 2) material and particle size of the aerosol are not important for the aerosol decay rate, 3) the aerosol decay rate is related to the decay heat production which is known function of time, and the relevant part of it must be assessed usually for other purposes as well. (orig.) [de

Temporal trends in human vulnerability to excessive heat

Science.gov (United States)

Sheridan, Scott C.; Allen, Michael J.

2018-04-01

Over recent decades, studies have examined various morbidity and mortality outcomes associated with heat exposure. This review explores the collective knowledge of the temporal trends of heat on human health, with regard to the hypothesis that humans are less vulnerable to heat events presently than in the past. Using Web of Science and Scopus, the authors identified all peer-reviewed articles that contained keywords on human impact (e.g. mortality, morbidity) and meteorological component (e.g. heat, heatwave). After sorting, a total of 71 articles, both case studies and epidemiological studies, contained explicit assessments of temporal trends in human vulnerability, and thus were used in this review. Most of the studies utilized mortality data, focused on the developed world, and showed a general decrease in heat sensitivity. Factors such as the implementation of a heat warning system, increased awareness, and improved quality of life were cited as contributing factors that led to the decreased impact of heat. Despite the overall recent decreases in heat vulnerability, spatial variability was shown, and differences with respect to health outcomes were also discussed. Several papers noted increases in heat’s impact on human health, particularly when unprecedented conditions occurred. Further, many populations, from outdoor workers to rural residents, in addition to the populations in much of the developing world, have been significantly underrepresented in research to date, and temporal changes in their vulnerability should be assessed in future studies. Moreover, continued monitoring and improvement of heat intervention is needed; with projected changes in the frequency, duration, and intensity of heat events combined with shifts in demographics, heat will remain a major public health issue moving forward.

MELCOR 1.8.2 Assessment: IET direct containment heating tests

Energy Technology Data Exchange (ETDEWEB)

Kmetyk, L.N.

1993-10-01

MELCOR is a fully integrated, engineering-level computer code, being developed at Sandia National Laboratories for the USNRC, that models the entire spectrum of severe accident phenomena in a unified framework for both BWRs and PWRS. As part of an ongoing assessment program, the MELCOR computer code has been used to analyze several of the IET direct containment heating experiments done at 1:10 linear scale in the Surtsey test facility at Sandia and at 1:40 linear scale in the corium-water thermal interactions (CWTI) COREXIT test facility at Argonne National Laboratory. These MELCOR calculations were done as an open post-test study, with both the experimental data and CONTAIN results available to guide the selection of code input. Basecase MELCOR results are compared to test data in order to evaluate the new HPME DCH model recently added in MELCOR version 1.8.2. The effect of various user-input parameters in the HPME model, which define both the initial debris source and the subsequent debris interaction, were investigated in sensitivity studies. In addition, several other non-default input modelling changes involving other MELCOR code packages were required in our IET assessment analyses in order to reproduce the observed experiment behavior. Several calculations were done to identify whether any numeric effects exist in our DCH IET assessment analyses.

THERMAL HYDRAULIC ISSUES OF CONTAINMENT FILTERED VENTING SYSTEM FOR A LONG OPERATING TIME

Directory of Open Access Journals (Sweden)

YOUNG SU NA

2014-12-01

Full Text Available This study investigated the thermal hydraulic issues in the Containment Filtered Venting System (CFVS for a long operating time using the MELCOR computer code. The modeling of the CFVS, including the models for pool scrubbing and the filter, was added to the input file for the OPR-1000, and a Station Blackout (SBO was chosen as an accident scenario. Although depressurization in the containment building as a primary objective of the CFVS was successful, the decontamination feature by scrubbing and filtering in the CFVS for a long operating time could fail by the continuous evaporation of the scrubbing solution. After the operation of the CFVS, the atmosphere temperature in the CFVS became slightly above the water saturation temperature owing to the release of an amount of steam with high temperature from the containment building to the scrubbing solution. Reduced pipe diameters at the inlet and outlet of the CFVS vessel mitigated the evaporation of scrubbing water by controlling the amount of high-temperature steam and the water saturation temperature.

Heat transfer modelling in a spent-fuel dry storage system

International Nuclear Information System (INIS)

Ritz, J.B.; Le Bonhomme, S.

2001-01-01

The purpose of this paper is to present a numerical modelling of heat transfers in a Spent-Fuel horizontal dry storage. The horizontal dry storage is an interesting issue to momentary store spent fuel containers before the final storage. From a thermal point of view, the cooling of spent fuel container by natural convection is a suitable and inexpensive process but it necessitates to well define the dimensions of the concept due to the difficulty to control the thermal environment. (author)

Heat pipe

International Nuclear Information System (INIS)

Triggs, G.W.; Lightowlers, R.J.; Robinson, D.; Rice, G.

1986-01-01

A heat pipe for use in stabilising a specimen container for irradiation of specimens at substantially constant temperature within a liquid metal cooled fast reactor, comprises an evaporator section, a condenser section, an adiabatic section therebetween, and a gas reservoir, and contains a vapourisable substance such as sodium. The heat pipe further includes a three layer wick structure comprising an outer relatively fine mesh layer, a coarse intermediate layer and a fine mesh inner layer for promoting unimpeded return of condensate to the evaporation section of the heat pipe while enhancing heat transfer with the heat pipe wall and reducing entrainment of the condensate by the upwardly rising vapour. (author)

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

Cooling device for reactor container

International Nuclear Information System (INIS)

Akiba, Miyuki.

1996-01-01

In a cooling device for a reactor container, a low pressure vessel is connected to an incondensible gas vent tube by way of an opening/closing valve. Upon occurrence of a loss of coolant accident, among steams and incondensible gases contained in the reactor container, steams are cooled and condensed in a heat exchanger. The incondensible gases are at first discharged from the heat exchanger to a suppression pool by way of the incondensible gas vent tube, but subsequently, they are stagnated in the incondensible gas vent tube to hinder heat exchanging and steam cooling and condensing effects in the heat exchanger thereby raising temperature and pressure in the reactor. However, if the opening/closing valve is opened when the incondensible gases are stagnated in the incondensible gas vent tube, since the incondensible gases stagnated in the heat exchanger are sucked and discharged to the low pressure vessel, the performance of the heat exchanger is maintained satisfactorily thereby enabling to suppress elevation of temperature and pressure in the reactor container. (N.H.)

Hydrogen related safety issues in the context of containments of Indian PHWRs

International Nuclear Information System (INIS)

Markendeya, S.G.; Ghosh, A.K.; Kushwaha, H.S.; Venkat Raj, V.

2002-01-01

Full text: Assessment of risk due to hydrogen released during postulated hypothetical accident scenarios in the nuclear power plants (NPPs) has been an important area of R and D studies world over. The issues, such as appropriate methodologies for estimation of hydrogen source term and for hydrogen dispersion calculations, technology development for hydrogen mitigation in containment of NPPs and assessment of damage due to deflagration/detonation of hydrogen (if it occurs) are being addressed as a part of some of the multidisciplinary study programs currently underway in BARC. While a significant overall progress has been achieved in general as a result of these programs, requirements of further fine-tuning of these studies have also emerged. The present paper takes a brief look at the current state-of the-art technology available to address these issues. The progress of R and D studies underway at BARC has also been critically reviewed in the paper to bring out necessary planning of further studies so as to enhance the safety of Indian NPPs

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.)

Soybean-derived biofuels and home heating fuels.

Science.gov (United States)

Mushrush, George W; Wynne, James H; Willauer, Heather D; Lloyd, Christopher L

2006-01-01

It is environmentally enticing to consider replacing or blending petroleum derived heating fuels with biofuels for many reasons. Major considerations include the soaring worldwide price of petroleum products, especially home heating oil, the toxicity of the petroleum-derived fuels and the environmental damage that leaking petroleum tanks afford. For these reasons, it has been suggested that domestic renewable energy sources be considered as replacements, or at the least, as blending stocks for home heating fuels. If recycled soy restaurant cooking oils could be employed for this purpose, this would represent an environmental advantage. Renewable plant sources of energy tend to be less toxic than their petroleum counterparts. This is an important consideration when tank leakage occurs. Home fuel oil storage tanks practically always contain some bottom water. This water environment has a pH value that factors into heating fuel stability. Therefore, the question is: would the biofuel help or exacerbate fuel stability and furnace maintenance issues?

Flammable gas issues in double-contained receiver tanks. Revision 2

International Nuclear Information System (INIS)

Peurrung, L.M.; Mahoney, L.A.; Stewart, C.W.; Gauglitz, P.A.; Pederson, L.R.; Bryan, S.A.; Shepard, C.L.

1998-08-01

Four double-contained receiver tanks (DCRTs) at Hanford will be used to store salt-well pumped liquids from tanks on the Flammable Gas Watch List. This document was created to serve as a reference document describing the current knowledge of flammable gas issues in DCRTs. The document identifies, describes, evaluates, and attempts to quantify potential gas carryover and release mechanisms. It estimates several key parameters needed for these calculations, such as initial aqueous concentrations and ventilation rate, and evaluates the uncertainty in those estimates. It justifies the use of the Schumpe model for estimating vapor-liquid equilibrium constants. It identifies several potential waste compatibility issues (such as mixing and pH or temperature changes) that could lead to gas release and provides a basis for calculating their effects. It evaluates the potential for gas retention in precipitated solids within a DCRT and whether retention could lead to a buoyant displacement instability (rollover) event. It discusses rates of radiolytic, thermal, and corrosive hydrogen generation within the DCRT. It also describes in detail the accepted method of calculating the lower flammability limit (LFL) for mixtures of flammable gases. The report incorporates these analyses into two models for calculating headspace flammability, one based on instantaneous equilibrium between dissolved gases and the headspace and one incorporating limited release rates based on mass-transfer considerations. Finally, it demonstrates the use of both models to estimate headspace flammable gas concentrations and minimum ventilation rates required to maintain concentrations below 25% of the LFL

Horizontal Heat Exchanger Design and Analysis for Passive Heat Removal Systems

Energy Technology Data Exchange (ETDEWEB)

Vierow, Karen

2005-08-29

This report describes a three-year project to investigate the major factors of horizontal heat exchanger performance in passive containment heat removal from a light water reactor following a design basis accident LOCA (Loss of Coolant Accident). The heat exchanger studied in this work may be used in advanced and innovative reactors, in which passive heat removal systems are adopted to improve safety and reliability The application of horizontal tube-bundle condensers to passive containment heat removal is new. In order to show the feasibility of horizontal heat exchangers for passive containment cooling, the following aspects were investigated: 1. the condensation heat transfer characteristics when the incoming fluid contains noncondensable gases 2. the effectiveness of condensate draining in the horizontal orientation 3. the conditions that may lead to unstable condenser operation or highly degraded performance 4. multi-tube behavior with the associated secondary-side effects This project consisted of two experimental investigations and analytical model development for incorporation into industry safety codes such as TRAC and RELAP. A physical understanding of the flow and heat transfer phenomena was obtained and reflected in the analysis models. Two gradute students (one funded by the program) and seven undergraduate students obtained research experience as a part of this program.

Advances on the time differential three-phase-lag heat conduction model and major open issues

Science.gov (United States)

D'Apice, Ciro; Zampoli, Vittorio

2017-07-01

The main purpose of this short contribution is to summarize the recent achievements concerning the so-called time differential three-phase-lag heat conduction model, contextually focusing attention on some of the numerous open problems associated with such an attractive theory. After having briefly recalled the origin of the model at issue, the restrictions upon the delay times and the constitutive tensors able to make it thermodynamically consistent are recalled. Under these hypotheses, the investigation of the well-posedness issue has already provided important results in terms of uniqueness and continuous dependence of the solutions (even related to the thermoelastic case), as well as in terms of existence of a domain of influence of the assigned data in connection with the thermoelastic model. Finally, some of the main problems currently object of investigation are recalled, including the very challenging issues about the different possible choices of Taylor series expansion orders for the constitutive equation, the interaction of the model with energy processes that take place on the nanoscale, with multi-porous materials and with biological systems.

Radioactive materials transporting container and vehicles

International Nuclear Information System (INIS)

Reese, S.L.

1980-01-01

A container and vehicle therefor for transporting radioactive materials is provided. The container utilizes a removable system of heat conducting fins made of a light weight highly heat conductive metal, such as aluminum or aluminum alloys. This permits a substantial reduction in the weight of the container during transport, increases the heat dissipation capability of the container and substantially reduces the scrubbing operation after loading and before unloading the radioactive material from the container. The vehicle utilizes only a pair of horizontal side beams interconnecting a pair of yoke members to support the container and provide the necessary strength and safety with a minimum of weight

Application of heat pipes in nuclear reactors for passive heat removal

Energy Technology Data Exchange (ETDEWEB)

Haque, Z.; Yetisir, M., E-mail: haquez@aecl.ca [Atomic Energy of Canada Limited, Chalk River, Ontario (Canada)

2013-07-01

This paper introduces a number of potential heat pipe applications in passive (i.e., not requiring external power) nuclear reactor heat removal. Heat pipes are particularly suitable for small reactors as the demand for heat removal is significantly less than commercial nuclear power plants, and passive and reliable heat removal is required. The use of heat pipes has been proposed in many small reactor designs for passive heat removal from the reactor core. This paper presents the application of heat pipes in AECL's Nuclear Battery design, a small reactor concept developed by AECL. Other potential applications of heat pipes include transferring excess heat from containment to the atmosphere by integrating low-temperature heat pipes into the containment building (to ensure long-term cooling following a station blackout), and passively cooling spent fuel bays. (author)

Facility for the storage of spent, heat-emitting and container-enclosed nuclear reactor fuel assemblies

International Nuclear Information System (INIS)

Hennings, U.

1987-01-01

Patent for facility for the storage of spent, heat-emitting and container-enclosed nuclear reactor fuel assemblies, which are arranged within a building in a horizontal position and are cooled by a gas stream, whereby the building has a storage and a loading zone, characterized by the fact that pallet trucks arranged one above the other in a row and such that an interspace is left for the receiving positions for the containers, the the pallet trucks can be moved along rails that extend between two side walls arranged opposite to one another in the storage zone, that the storage zone can be loaded and unloaded by opening located in these two side walls, and that the gas stream only circulates within the building

Magnetic fusion energy plasma interactive and high heat flux components. Volume II. Technical assessment of the critical issues and problem areas in high heat flux materials and component development

International Nuclear Information System (INIS)

Abdou, M.A.; Boyd, R.D.; Easor, J.R.

1984-06-01

A technical assessment of the critical issues and problem areas for high heat flux materials and components (HHFMC) in magnetic fusion devices shows these problems to be of critical importance for the successful operation of near-term fusion experiments and for the feasibility and attractiveness of long-term fusion reactors. A number of subgroups were formed to assess the critical HHFMC issues along the following major lines: (1) source conditions, (2) systems integration, (3) materials and processes, (4) thermal hydraulics, (5) thermomechanical response, (6) electromagnetic response, (7) instrumentation and control, and (8) test facilities. The details of the technical assessment are presented in eight chapters. The primary technical issues and needs for each area are highlighted

Magnetic fusion energy plasma interactive and high heat flux components. Volume II. Technical assessment of the critical issues and problem areas in high heat flux materials and component development

Energy Technology Data Exchange (ETDEWEB)

Abdou, M.A.; Boyd, R.D.; Easor, J.R.; Gauster, W.B.; Gordon, J.D.; Mattas, R.F.; Morgan, G.D.; Ulrickson, M.A,; Watson, R.D.; Wolfer, W.G,

1984-06-01

A technical assessment of the critical issues and problem areas for high heat flux materials and components (HHFMC) in magnetic fusion devices shows these problems to be of critical importance for the successful operation of near-term fusion experiments and for the feasibility and attractiveness of long-term fusion reactors. A number of subgroups were formed to assess the critical HHFMC issues along the following major lines: (1) source conditions, (2) systems integration, (3) materials and processes, (4) thermal hydraulics, (5) thermomechanical response, (6) electromagnetic response, (7) instrumentation and control, and (8) test facilities. The details of the technical assessment are presented in eight chapters. The primary technical issues and needs for each area are highlighted.

Selection of cooling fluid for an organic Rankine cycle unit recovering heat on a container ship sailing in the Arctic region

DEFF Research Database (Denmark)

Suárez de la Fuente, Santiago; Larsen, Ulrik; Pierobon, Leonardo

2017-01-01

As Arctic sea ice coverage declines it is expected that marine traffic could increase in this northern region due to shorter routes. Navigating in the Arctic offers opportunities and challenges for waste heat recovery systems (WHRS). Lower temperatures require larger heating power on board, hence...... air as coolant. This paper explores the use of two different coolants, air and seawater, for an organic Rankine cycle (ORC) unit using the available waste heat in the scavenge air system of a container ship navigating in Arctic Circle. Using a two-step single objective optimisation process, detailed...

Technologies and Materials for Recovering Waste Heat in Harsh Environments

Energy Technology Data Exchange (ETDEWEB)

Nimbalkar, Sachin U. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Thekdi, Arvind [E3M, Inc. North Potomac, MD (United States); Rogers, Benjamin M. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Kafka, Orion L. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Wenning, Thomas J. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

2014-12-15

A large amount (7,204 TBtu/year) of energy is used for process heating by the manufacturing sector in the United States (US). This energy is in the form of fuels mostly natural gas with some coal or other fuels and steam generated using fuels such as natural gas, coal, by-product fuels, and some others. Combustion of these fuels results in the release of heat, which is used for process heating, and in the generation of combustion products that are discharged from the heating system. All major US industries use heating equipment such as furnaces, ovens, heaters, kilns, and dryers. The hot exhaust gases from this equipment, after providing the necessary process heat, are discharged into the atmosphere through stacks. This report deals with identification of industries and industrial heating processes in which the exhaust gases are at high temperature (>1200 F), contain all of the types of reactive constituents described, and can be considered as harsh or contaminated. It also identifies specific issues related to WHR for each of these processes or waste heat streams.

The buffer/container experiment: results, synthesis, issues

International Nuclear Information System (INIS)

Graham, J.; Chandler, N.A.; Dixon, D.A.; Roach, P.J.; To, T.; Wan, A.W.L.

1997-12-01

A large in-ground experiment has examined how heat affects the performance of the dense sand bentonite 'buffer' that has been proposed for use in the Canadian Nuclear Fuel Waste Management Program. The experiment was performed by Atomic Energy of Canada Limited at its Underground Research Laboratory, Lac du Bonnet, Manitoba between 1991 and 1994. The experiment placed a full-size heater representing a container of nuclear fuel waste in a 1.24-m diameter borehole filled with buffer below the floor of a room excavated at 240-m depth in granitic rock of the Canadian Shield. The buffer and surrounding rock were extensively instrumented for temperatures, total pressures, water pressures, suctions, and rock displacements. Power was provided to the heater for almost 900 days. The experiment showed that good rock conditions can be pre-selected, a borehole can be drilled, and buffer can be placed at controlled densities and water contents. The instrumentation generally worked well, and an extensive data base was successfully organized. Drying was observed in buffer close to the heater. This caused some desiccation cracking. However the cracks only extended approximately one third of the distance to the buffer-rock interface and did not form an advective pathway. Following sampling at the time of decommissioning, cracked samples of buffer were transported to the laboratory and given access to water. The hydraulic conductivities and swelling pressures of these resaturated samples were very similar to those of uncracked buffer. A good balance was achieved between the mass of water flowing into the experiment from the surrounding rock and the increased mass of water in the buffer. A good understanding was developed of the relationships between suctions, water contents, and total pressures in buffer near the buffer-rock interface. Comparisons between measurements and predictions of measured parameters show that a good understanding has been developed of the processes operating

The buffer/container experiment: results, synthesis, issues

Energy Technology Data Exchange (ETDEWEB)

Graham, J. [Univ. of Manitoba, Dept. of Civil Engineering, Winnipeg, MB (Canada); Chandler, N.A.; Dixon, D.A.; Roach, P.J.; To, T.; Wan, A.W.L

1997-12-01

A large in-ground experiment has examined how heat affects the performance of the dense sand bentonite 'buffer' that has been proposed for use in the Canadian Nuclear Fuel Waste Management Program. The experiment was performed by Atomic Energy of Canada Limited at its Underground Research Laboratory, Lac du Bonnet, Manitoba between 1991 and 1994. The experiment placed a full-size heater representing a container of nuclear fuel waste in a 1.24-m diameter borehole filled with buffer below the floor of a room excavated at 240-m depth in granitic rock of the Canadian Shield. The buffer and surrounding rock were extensively instrumented for temperatures, total pressures, water pressures, suctions, and rock displacements. Power was provided to the heater for almost 900 days. The experiment showed that good rock conditions can be pre-selected, a borehole can be drilled, and buffer can be placed at controlled densities and water contents. The instrumentation generally worked well, and an extensive data base was successfully organized. Drying was observed in buffer close to the heater. This caused some desiccation cracking. However the cracks only extended approximately one third of the distance to the buffer-rock interface and did not form an advective pathway. Following sampling at the time of decommissioning, cracked samples of buffer were transported to the laboratory and given access to water. The hydraulic conductivities and swelling pressures of these resaturated samples were very similar to those of uncracked buffer. A good balance was achieved between the mass of water flowing into the experiment from the surrounding rock and the increased mass of water in the buffer. A good understanding was developed of the relationships between suctions, water contents, and total pressures in buffer near the buffer-rock interface. Comparisons between measurements and predictions of measured parameters show that a good understanding has been developed of the processes

Analysis of direct containment heating in Ling'ao-2 nuclear power plant model With FCI-CMFD software

International Nuclear Information System (INIS)

Wang Xi; Yang Yanhua; Huang Xi

2010-01-01

Ling'ao 2 Nuclear Power Plant (NPP) model was established in the paper. Three dimensional numerical calculation code MC3D was used for analyzing the potential direct containment heating (DCH) accident in the model. MC3D is a special software for fuel coolant interaction (FCI) analysis. In order to predict the accident accurately, initial conditions provided by the SBO accident calculation and the geometric model of Ling'ao 2 NPP were combined to simulate the process of the accident. The calculation gives temperature field, droplet Janume fraction field, velocity field, and the pressure variation in the containment in the case of accident. The result shows that DCH will cause the pressure rising rapidly in the containment and in some local areas temperature rises too. (authors)

Controlling the heating mode of heat pumps with the TRIANA three step methodology

NARCIS (Netherlands)

Toersche, Hermen; Bakker, Vincent; Molderink, Albert; Nykamp, Stefan; Hurink, Johann L.; Smit, Gerardus Johannes Maria

2012-01-01

Heat pump based heating systems are increasingly becoming an economic and efficient alternative for domestic gas heating systems. Concentrations of heat pump installations do consume large amounts of electricity, causing significant grid distribution and stability issues when the diversity factor is

Hydro-structural issues in the design of ultra large container ships

Directory of Open Access Journals (Sweden)

Sime Malenica

2014-12-01

Full Text Available The structural design of the ships includes two main issues which should be checked carefully, namely the extreme structural response (yielding & buckling and the fatigue structural response. Even if the corresponding failure modes are fundamentally different, the overall methodologies for their evaluation have many common points. Both issues require application of two main steps: deterministic calculations of hydro-structure interactions for given operating conditions on one side and the statistical post-processing in order to take into account the lifetime operational profile, on the other side. In the case of ultra large ships such as the container ships and in addition to the classical quasi-static type of structural responses the hydroelastic structural response becomes important. This is due to several reasons among which the following are the most important: the increase of the flexibility due to their large dimensions (Lpp close to 400 m which leads to the lower structural natural frequencies, very large operational speed (20 knots and large bow flare (increased slamming loads. The correct modeling of the hydroelastic ship structural response, and its inclusion into the overall design procedure, is significantly more complex than the evaluation of the quasi static structural response. The present paper gives an overview of the different tools and methods which are used in nowadays practice.

Thulium-170 heat source

Science.gov (United States)

Walter, Carl E.; Van Konynenburg, Richard; VanSant, James H.

1992-01-01

An isotopic heat source is formed using stacks of thin individual layers of a refractory isotopic fuel, preferably thulium oxide, alternating with layers of a low atomic weight diluent, preferably graphite. The graphite serves several functions: to act as a moderator during neutron irradiation, to minimize bremsstrahlung radiation, and to facilitate heat transfer. The fuel stacks are inserted into a heat block, which is encased in a sealed, insulated and shielded structural container. Heat pipes are inserted in the heat block and contain a working fluid. The heat pipe working fluid transfers heat from the heat block to a heat exchanger for power conversion. Single phase gas pressure controls the flow of the working fluid for maximum heat exchange and to provide passive cooling.

The Economics of Connecting of Small Buildings to Geothermal District Heating Systems

Energy Technology Data Exchange (ETDEWEB)

Rafferty, Kevin

2003-03-01

Many of the communities co-located with geothermal resources are very small and as a result the buildings they contain tend to be small as well. Generally, small buildings (10,000 ft2) use heating systems which are not hot water based. Since geothermal district heating systems deliver hot water, the costs associated with the conversion of small building heating systems to use hot water for heating is an issue of great influence in terms of the potential development of such systems. This paper examines the typical retrofit costs associated with conversion of small buildings and the level of savings necessary to attract the interest of owners. In general, the prospects for connection of such buildings based only on energy savings is not positive.

The economics of connecting of small buildings to geothermal district heating systems

Energy Technology Data Exchange (ETDEWEB)

Rafferty, Kevin

2001-01-01

Many of the communities co-located with geothermal resources are very small and as a result the buildings they contain tend to be small as well. Generally, small buildings (10,000 ft2) use heating systems which are not hot water based. Since geothermal district heating systems deliver hot water, the costs associated with the conversion of small building heating systems to use hot water for heating is an issue of great influence in terms of the potential development of such systems. This paper examines the typical retrofit costs associated with conversion of small buildings and the level of savings necessary to attract the interest of owners. In general, the prospects for connection of such buildings based only on energy savings is not positive.

Heat recovery apparatus

International Nuclear Information System (INIS)

McFarland, I.

1987-01-01

Heat transfer is a living science and technical advances are constantly being made. However, in many cases, progress is limited by the equipment that is available on the market, rather than by knowledge of the heat transfer process. A case in point is the design of economizers: in such equipment a small quantity of water (with a relatively good heat transfer coefficient) is heated by a large quantity of low-pressure gas (with an inherently low heat transfer coefficient). As a first step in design finned tubing is used to lessen the discrepancy in coefficients. From this point, it becomes apparent that the equipment consists of a small number of tubes (to maintain good velocity on the water side) of considerable length (to provide sufficient area). In the process industries the base pressure, though low, may be in the region of 0.5 bar, and there is no convenient flue in which to place the heat recovery coil. It is therefore contained in a flat-sided enclosure, which is ill-fitted to pressure containment and is therefore reinforced with a plethora of structural sections. Such inelegant construction is quite common in North America; in Europe, cylindrical containments of vast size have been supplied for the same purposes. The real shortcoming is a successful marriage of different disciplines to produce reliable and efficient heat transfer equipment suitably contained

Renewable heat: Waste heat, an emerging sector full of resources. An eco-district fed by the heat from a data center; Integrating objectives matching the ambitions, clear and on the long term; High-energy wastes in Brittany: Brest is securing its heat network with multiple energies

International Nuclear Information System (INIS)

Richard, Aude

2017-01-01

This file on renewable heat contains four articles which themes are: waste heat (from incineration plants or industrial sites) is an emerging sector full of resources, which benefits now of a financial subsidy from ADEME, the French organization for energy and environment; an example is given with Chamtor, a highly energy-consuming cereal transformer. The second article presents an urban eco-district in Paris that is fed by the heat issued by a data center. The third article (Integrating objectives matching the ambitions, clear and on the long term) presents some recommendations from the French Renewable Energy Association (SER) towards a better energy valorization of residual wastes. The fourth article presents two examples of energy valorization of wastes in Brittany, one with the valorization of high-energy solid wastes for supplying heat to a milk farm and greenhouses, the other one concerns the development of a heat network supplied by an incineration plant, with a mix of energy sources such as wastes, wood and gas, and a special juridical scheme

Comparing the analgesic effect of heat patch containing iron chip and ibuprofen for primary dysmenorrhea: a randomized controlled trial

Directory of Open Access Journals (Sweden)

Navvabi Rigi Shahindokht

2012-08-01

Full Text Available Abstract Background Primary dysmenorrhea is a common and sometimes disabling condition. In recent years, some studies aimed to improve the treatment of dysmenorrhea, and therefore, introduced several therapeutic measures. This study was designed to compare the analgesic effect of iron chip containing heat wrap with ibuprofen for the treatment of primary dysmenorrhea. Methods In this randomized (IRCT201107187038N2 controlled trial, 147 students (18–30 years old with the diagnosis of primary dysmenorrhea were enrolled considering the CONSORT guideline. Screening for primary dysmenorrhea was done by a two-question screening tool. The participants were randomly assigned into one of the intervention groups (heat Patch and ibuprofen. Data regarding the severity and emotional impact of the pain were recorded by a shortened version of McGill Pain Questionnaire (SF-MPQ. Student's t test was used for statistical analysis. Results The maximum and minimum pain severities were observed at 2 and 24 hours in both groups. The severity of sensual pain at 8, 12, and 24 hours was non-significantly less in the heat Patch group. There was also no significant difference between the groups regarding the emotional impact of pain at the first 2, 4, 8, 12 and 12 hours of menstruation. Conclusions Heat patch containing Iron chip has comparable analgesic effects to ibuprofen and can possibly be used for primary dysmenorrhea. Trial registration IRCT201107187038N2

Interactive economic analysis of small-scale heating plant

Energy Technology Data Exchange (ETDEWEB)

Landen, R.A.; Sanders, D.L.; Douglas, B.H.

1998-11-01

This report contains the work that has been undertaken by LRZ Limited in pursuance of this agreement. The potential for small-scale biomass heating systems is identified, and surrounding issues relating to acceptance are discussed. Such systems are described, and the origins of capital and running costs examined. A full review of the calculation methods for boiler plant size and fuel consumption is made, and subsequently expounded in four varying case studies. The results of this work are discussed, and the final development of the computer models is reviewed, incorporating further refinements to the method. Finally, data not contained in the text is incorporated in comprehensive appendices. (author)

Numerical Heat Transfer Studies of a Latent Heat Storage System Containing Nano-Enhanced Phase Change Material

Directory of Open Access Journals (Sweden)

S F Hosseinizadeh

2011-01-01

Full Text Available The heat transfer enhancement in the latent heat thermal energy storage system through dispersion of nanoparticle is reported. The resulting nanoparticle-enhanced phase change materials (NEPCM exhibit enhanced thermal conductivity in comparison to the base material. The effects of nanoparticle volume fraction and some other parameters such as natural convection are studied in terms of solid fraction and the shape of the solid-liquid phase front. It has been found that higher nanoparticle volume fraction result in a larger solid fraction. The present results illustrate that the suspended nanoparticles substantially increase the heat transfer rate and also the nanofluid heat transfer rate increases with an increase in the nanoparticles volume fraction. The increase of the heat release rate of the NEPCM shows its great potential for diverse thermal energy storage application.

Water spray interaction with air-steam mixtures under containment spray conditions: comparison of heat and mass transfer modelling with the TOSQAN spray tests

International Nuclear Information System (INIS)

Malet, J.; Lemaitre, P.; Porcheron, E.; Vendel, J.

2005-01-01

Full text of publication follows: During the course of a hypothetical severe accident in a Pressurized Water Reactor (PWR), hydrogen can be produced by the reactor core oxidation and distributed into the reactor containment according to convection flows and water steam wall condensation. In order to mitigate the risk of detonation generated by a high local hydrogen concentration, spray systems are used in the containment. The TOSQAN programme has been created to simulate separate-effect tests representative of typical accidental thermal-hydraulic flow conditions in the reactor containment. The present work concerns the interaction of a water spray, used at the top of the containment in order to reduce the steam partial pressure, with air-steam mixtures. The main phenomena occurring when water spray is used are the mixing induced by spray entrainment and the condensation on droplets. In order to improve the latter phenomena, different levels of modelling can be used. The objective of this paper is to analyze experimental results obtained for water spray interaction with air-steam mixtures using different heat and mass transfer modelling. For this purpose, two modelling issues have been used: the first one is devoted for the determination of the gas thermodynamical properties, and the second one concerns the droplets characterization. In the first one, the gas thermodynamical analysis is performed using depressurization, gas temperature variation and humidity decrease during the spray injection. In this modelling, heat and mass transfer between the spray and the surrounding gas is treated in a global way by energy balance between the total amount of water and the gas. In the second one, droplets characterization is obtained by means of droplet size, temperature and velocities evolutions. In this modelling, the spray is considered as a single droplet falling with an initial velocity. Droplet interactions are neglected. Assessment of these two modelling is performed

Helical type vacuum container

International Nuclear Information System (INIS)

Owada, Kimio.

1989-01-01

Helical type vacuum containers in the prior art lack in considerations for thermal expansion stresses to helical coils, and there is a possibility of coil ruptures. The object of the present invention is to avoid the rupture of helical coils wound around the outer surface of a vacuum container against heat expansion if any. That is, bellows or heat expansion absorbing means are disposed to a cross section of a helical type vacuum container. With such a constitution, thermal expansion of helical coils per se due to temperature elevation of the coils during electric supply can be absorbed by expansion of the bellows or absorption of the heat expansion absorbing means. Further, this can be attained by arranging shear pins in the direction perpendicular to the bellows axis so that the bellows are not distorted when the helical coils are wound around the helical type vacuum container. (I.S.)

Heat transfer in a compact heat exchanger containing rectangular channels and using helium gas

Science.gov (United States)

Olson, D. A.

1991-01-01

Development of a National Aerospace Plane (NASP), which will fly at hypersonic speeds, require novel cooling techniques to manage the anticipated high heat fluxes on various components. A compact heat exchanger was constructed consisting of 12 parallel, rectangular channels in a flat piece of commercially pure nickel. The channel specimen was radiatively heated on the top side at heat fluxes of up to 77 W/sq cm, insulated on the back side, and cooled with helium gas flowing in the channels at 3.5 to 7.0 MPa and Reynolds numbers of 1400 to 28,000. The measured friction factor was lower than that of the accepted correlation for fully developed turbulent flow, although the uncertainty was high due to uncertainty in the channel height and a high ratio of dynamic pressure to pressure drop. The measured Nusselt number, when modified to account for differences in fluid properties between the wall and the cooling fluid, agreed with past correlations for fully developed turbulent flow in channels. Flow nonuniformity from channel-to-channel was as high as 12 pct above and 19 pct below the mean flow.

Joule-Heated Molten Regolith Electrolysis Reactor Concepts for Oxygen and Metals Production on the Moon and Mars

Science.gov (United States)

Sibille, Laurent; Dominguez, Jesus A.

2012-01-01

The technology of direct electrolysis of molten lunar regolith to produce oxygen and molten metal alloys has progressed greatly in the last few years. The development of long-lasting inert anodes and cathode designs as well as techniques for the removal of molten products from the reactor has been demonstrated. The containment of chemically aggressive oxide and metal melts is very difficult at the operating temperatures ca. 1600 C. Containing the molten oxides in a regolith shell can solve this technical issue and can be achieved by designing a Joule-heated (sometimes called 'self-heating') reactor in which the electrolytic currents generate enough Joule heat to create a molten bath. Solutions obtained by multiphysics modeling allow the identification of the critical dimensions of concept reactors.

Method of making Tl-Sr-Ca-Cu-oxide superconductors comprising heating at elevated pressures in a sealed container

International Nuclear Information System (INIS)

Lechtev, W.L.; Osofsky, M.S.; Skelton, E.F.; Toth, L.E.

1992-01-01

This patent describes a method of forming a Tl-Sr-Ca-Cu-oxide high T c superconductor. It comprises forming a reaction mixture of the oxides of Sr, Cu, Ca, and Tl in stoichiometric proportions to make a Tl-Sr-Ca-Cu-oxide high T c superconducting compound; compressing the reaction mixture into a hard body; placing the hard body into a container for containing thallium vapor; evacuating and sealing the hard body in the container; heating the hard body and the container at a temperature of about 800 degrees C to about 950 degrees C and under pressure of at least about 30,000 psi until the container metal around the hard body and the oxides of Tl, Sr, Ca, and Cu react to form a superconducting compound; and cooling the superconducting compound to room temperature and returning the superconducting compound to atmospheric pressure

Alternative cooling water flow path for RHR heat exchanger and its effect on containment response during extended station blackout for Chinshan BWR-4 plant

Energy Technology Data Exchange (ETDEWEB)

Yuann, Yng-Ruey, E-mail: ryyuann@iner.gov.tw

2016-04-15

Highlights: • Motivating alternative RHR heat exchanger tube-side flow path and determining required capacity. • Calculate NSSS and containment response during 24-h SBO for Chinshan BWR-4 plant. • RETRAN and GOTHIC models are developed for NSSS and containment, respectively. • Safety relief valve blowdown flow and energy to drywell are generated by RETRAN. • Analyses are performed with and without reactor depressurization, respectively. - Abstract: The extended Station Blackout (SBO) of 24 h has been analyzed with respect to the containment response, in particular the suppression pool temperature response, for the Chinshan BWR-4 plant of MARK-I containment. The Chinshan plant, owned by Taiwan Power Company, has twin units with rated core thermal power of 1840 MW each. The analysis is aimed at determining the required alternative cooling water flow capacity for the residual heat removal (RHR) heat exchanger when its tube-side sea water cooling flow path is blocked, due to some reason such as earthquake or tsunami, and is switched to the alternative raw water source. Energy will be dissipated to the suppression pool through safety relief valves (SRVs) of the main steam lines during SBO. The RETRAN model is used to calculate the Nuclear Steam Supply System (NSSS) response and generate the SRV blowdown conditions, including SRV pressure, enthalpy, and mass flow rate. These conditions are then used as the time-dependent boundary conditions for the GOTHIC code to calculate the containment pressure and temperature response. The shaft seals of the two recirculation pumps are conservatively assumed to fail due to loss of seal cooling and a total leakage flow rate of 36 gpm to the drywell is included in the GOTHIC model. Based on the given SRV blowdown conditions, the GOTHIC containment calculation is performed several times, through the adjustment of the heat transfer rate of the RHR heat exchanger, until the criterion that the maximum suppression pool temperature

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

Conceptual design for Japan Sodium-Cooled Fast Reactor. (4) Developmental study of steel plate reinforced concrete containment vessel for JSFR

International Nuclear Information System (INIS)

Hosoya, Takusaburo; Negishi, Kazuo; Satoh, Kenichiro; Somaki, Takahiro; Matsuo, Ippei; Shimizu, Katsusuke

2009-01-01

An innovative containment vessel, namely Steel plate reinforced Concrete Containment Vessel (SCCV) is developed for Japan Sodium-Cooled Fast Reactor (JSFR). Reducing plant construction cost is one of the most important issues for commercialization of fast reactors. This study investigated construction issues including the building structure and the construction method as well as design issues in terms of the applicability of SCCV to fast reactors. An experimental study including loading and/or heating tests has been carried out to investigate the fundamental structural features, which would be provided to develop methodology to evaluate the feasibility of SCCV under the severe conditions. In this paper, the test plan is described as well as the first test results. (author)

The effect of heat treatment and irradiation on some physical properties of lithium borate glass containing transition elements

International Nuclear Information System (INIS)

Soliman, A.A.; Aly, S.A.; Frhan, H.; Abo-Zeid, Y.M.

1999-01-01

The effect of introducing iron, nickel and cobalt oxide on some physical properties investigated in this article. The electrical conductivity has a higher value in samples containing 1 mol% transition metal oxides. The conductivity decreased as the content of transition metal oxide increased up to 5 mol% which was due to the change of Fe 2+ to Fe 3+ and increase of Co and Ni ions in octahedral state. The effect of heat treatment on the conductivity measurements shows a decrease in the conductivity values for glass samples with increasing the heat treatment time up to 72 h. This decrease could be attributed to the change in the structure of the glass samples. The investigation of radiation doses with the electrical conductivity concluded that the conductivity increased with increase the irradiation doses. The reason of that may be due to increasing the number of vacancies and vacancy interstitial pairs which are created. The magnetic susceptibility measurements showed an increase in the magnetic susceptibility as Fe 2 O 3 and NiO were increased. While for samples containing CoO the magnetic susceptibility changed due to the change in coordination number of the Co ions. The effect of heat treatment on magnetic susceptibility of the investigated samples concluded that the magnetic susceptibilities have a random behavior with increasing time of heat treatment. By investigating irradiation doses with a magnetic susceptibility it was found that the increase of irradiation dose promotes a tendency to change the magnetic susceptibility values. This change can be related to the presence of structure defects and impurities in the samples before irradiation

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

Reaction kinetics and reaction heat on thermal decomposition of solvent containing unstable reactive hydrocarbons with nitric acid at Tomsk-7 reprocessing plant

International Nuclear Information System (INIS)

Nishio, Gunji; Watanabe, Kouji; Koike, Tadao; Miyato, Teijiro.

1996-12-01

For analyzing a cause of the Tomsk-7 accident at Russian reprocessing plant, it is necessary to determine reaction-rate constant and reaction heat for a thermal decomposition of TBP/kerosine containing unstable reactive hydrocarbons with nitric acid. In JAERI, the rate constant and reaction heat were obtained from data measured with a differential thermal analyzer (DTA) for unstable hydrocarbons such as n-butanol, n-butyl nitrate, aromatic hydrocarbons, and cyclic compounds. The safety evaluation of Tomsk tank ruptured by the reaction was carried out by heat balance calculations between heat generation and heat loss in the tank using these rate constants and reaction heats. Consequently, it is clear that the cause of the tank rupture would be due to an exothermic reaction of aromatic hydrocarbons in kerosine made by petroleum with the concentrated nitric acid of 14.2N. (author)

Installation method for the steel container and vessel of the nuclear heating reactor

International Nuclear Information System (INIS)

Chen Liying; Guo Jilin; Liu Wei

2000-01-01

The Nuclear Heating Reactor (NHR) has the advantages of inherent safety and better economics, integrated arrangement, full power natural circulation and dual vessel structure. However, the large thin container presents a new and difficult problem. The characteristics of the dual vessel installation method are analyzed with system engineering theory. Since there is no foreign or domestic experience, a new method was developed for the dual vessel installation for the 5 MW NHR. The result shows that the installation method is safe and reliable. The research on the dual vessel installation method has important significance for the design, manufacture and installation of the NHR dual vessel, as well as the industrialization and standardization of the NHR

Study on the CFD simulation of refrigerated container

Science.gov (United States)

Arif Budiyanto, Muhammad; Shinoda, Takeshi; Nasruddin

2017-10-01

The objective this study is to performed Computational Fluid Dynamic (CFD) simulation of refrigerated container in the container port. Refrigerated container is a thermal cargo container constructed from an insulation wall to carry kind of perishable goods. CFD simulation was carried out use cross sectional of container walls to predict surface temperatures of refrigerated container and to estimate its cooling load. The simulation model is based on the solution of the partial differential equations governing the fluid flow and heat transfer processes. The physical model of heat-transfer processes considered in this simulation are consist of solar radiation from the sun, heat conduction on the container walls, heat convection on the container surfaces and thermal radiation among the solid surfaces. The validation of simulation model was assessed uses surface temperatures at center points on each container walls obtained from the measurement experimentation in the previous study. The results shows the surface temperatures of simulation model has good agreement with the measurement data on all container walls.

Direct containment heating experiments in Zion Nuclear Power Plant geometry using prototypic materials

International Nuclear Information System (INIS)

Binder, J.L.; McUmber, L.M.; Spencer, B.W.

1993-01-01

Direct Containment Heating (DCH) experiments have been completed which utilize prototypic core materials. The experiments reported on here are a continuation of the Integral Effects Testing (IET) DCH program. The experiments incorporated a 1/40 scale model of the Zion Nuclear Power Plant containment structures. The model included representations of the primary system volume, RPV lower head, cavity and instrument tunnel, and the lower containment structures. The experiments were steam driven. Iron-alumina thermite with chromium was used as a core melt stimulant in the earlier IET experiments. These earlier IET experiments at Argonne National Laboratory (ANL) and Sandia National Laboratories (SNL) provided useful data on the effect of scale on DCH phenomena; however, a significant question concerns the potential experiment distortions introduced by the use of non-prototypic iron/alumina thermite. Therefore, further testing with prototypic materials has been carried out at ANL. Three tests have been completed, DCH-U1A, U1B and U2. DCH-U1A and U1B employed an inerted containment atmosphere and are counterpart to the IET-1RR test with iron/alumina thermite. DCH-U2 employed nominally the same atmosphere composition of its counterpart iron/alumina test, IET-6. All tests, with prototypic material, have produced lower peak containment pressure rises; 45, 111 and 185 kPa in U1A, U1B and U2, compared to 150 and 250 kPa IET-1RR and 6. Hydrogen production, due to metal-steam reactions, was 33% larger in U1B and U2 compared to IET-1RR and IET-6. The pressurization efficiency was consistently lower for the corium tests compared to the IET tests

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)

Facility with a nuclear district heating reactor

International Nuclear Information System (INIS)

Straub, H.

1988-01-01

The district heating reactor has a pressure vessel which contains the reactor core and at least one coolant conducting primary heat carrier surrounded by a heat sink. The pressure vessel has two walls with a space between them. This space is connected with a container which contains air as heat isolating medium and water as heat conducting medium. During the normal reactor operation the space is filled by air from the container with the aid of a blower, whereas in the case of a break-down of the cooling system it is filled by water which flows out of the container by gravity after the blower has been switched off. The after-heat, generated in the reactor core during cooling break-down, is removed into the heat sink surrounding the pressure vessel in a safe and simple way. 6 figs

Effect of free swirl flow on the rate of mass and heat transfer at the bottom of a vertical cylindrical container and possible applications

International Nuclear Information System (INIS)

Konsowa, A.H.; Abdel-Aziz, M.H.; Abdo, M.S.E.; Hassan, M.S.; Sedahmed, G.H.

2017-01-01

Highlights: • Mass transfer at the bottom of a cylindrical container was studied under decaying swirl flow. • Parameters studied are swirl flow velocity, diameter of the inlet nozzle and solution properties. • A dimensionless equation was obtained using the significant parameters. • The present results were compared with the results obtained using perpendicular inlet nozzle. • Relevance of study to the design of membrane processes was highlighted. - Abstract: Rates of mass transfer at the base of a vertical cylindrical container were determined under decaying swirl flow by the electrochemical technique. Variables studied were swirl flow solution velocity, diameter of the tangential inlet nozzle and physical properties of the solution. The data were correlated by a dimensionless mass transfer equation. The equation can be used to predict the rate of heat loss from the bottom of swirl flow equipment as well as the rate of diffusion controlled corrosion of the bottom. The importance of the derived equation in the design and scale up of a cylindrical batch recirculating catalytic or electrochemical reactor with a catalyst layer or electrode at the bottom and a cooling jacket around the vertical wall suitable for conducting exothermic liquid – solid diffusion controlled reactions which need rapid temperature control to avoid the loss of heat sensitive catalysts or heat sensitive products was pointed out. Comparison of the present results with the results obtained using perpendicular inlet nozzle which generates parallel flow at the bottom and axial flow along the cylindrical container revealed the fact that although swirl flow produces higher rates of heat and mass transfer at the cylindrical wall than axial flow and the reverse is true at the container base. Relevance of the present study to the design and operation of membrane processes and heat recovery from hot pools of liquid metals and low melting alloys in the production stage was highlighted.

The Apatity nuclear heating plant project: modern technical and economic issues of nuclear heat application in Russia

International Nuclear Information System (INIS)

Adamov, E.O.; Romenkov, A.A.

1998-01-01

Traditionally Russia is a country with advanced structure of centralized heat supply. Many thermal plants and heating networks need technical upgrading to improve their technical and economic efficiency. Fossil fuelled heating capacities have a negative influence on ecology, which can be seen especially in the northern regions of Russia. Furthermore, fossil fuel prices are rising in Russia. The above factors tend to intensify the need for alternative heat sources being capable of solving the problem. Nuclear heat sources may be the alternative. In this paper, the main features of a proposed NHP in the Murmansk region are summarized. (author)

Study on Induction Heating Coil for Uniform Mold Cavity Surface Heating

OpenAIRE

Yu-Ting Sung; Sheng-Jye Hwang; Huei-Huang Lee; Durn-Yuan Huang

2014-01-01

Recently, energy saving is one of the important issues for polymer processing industry. Electromagnetic induction heating has many advantages such as fast heating and low energy consumption. Previous studies using electromagnetic induction heating for rapid tool heating have indicated that the temperature uniformity on a cavity surface is not easy to be achieved. In this paper, two different coils were used for heating uniform 7 mm thick hot work tool steel (JIS SKD61) surface. One is a four-...

Protected isotope heat source

International Nuclear Information System (INIS)

Burns, R.K.; Shure, L.I.; Katzen, E.D.

1975-01-01

A radioactive isotope capsule is disposed in a container (heat shield) which will have a single stable trim attitude when reentering the earth's atmosphere and while falling to earth. The center of gravity of the heat source is located forward of the midpoint between the front face and the rear face of the container. The capsule is insulated from the front face of the container but not from the rear surface of the container. (auth)

Steady natural convection in a horizontal channel containing heated rectangular blocks periodically mounted on its lower wall

International Nuclear Information System (INIS)

Bakkas, M.; Amahmid, A.; Hasnaoui, M.

2006-01-01

In this paper, we perform a numerical investigation of laminar steady natural convection flows in a two-dimensional horizontal channel containing heating rectangular blocks, periodically mounted on its lower wall. The blocks are heated at a constant temperature, T H ' and connected with adiabatic surfaces. The upper wall of the channel is maintained at a cold temperature T C ' . The parameters governing the problem are the Rayleigh number (10 2 = 6 ), the geometric parameter C (0.25=< C=l'/H'=<0.75) and the relative height of the blocks (1/8=< B=h'/H'=<1/2). The effect of the computational domain choice on the multiplicity of solutions is also investigated. The results obtained using air (Pr=0.72) as the working fluid show that the parameters B and C have a significant effect on the fluid flow and temperature fields. The symmetry of the flow is not always maintained although the boundary conditions for this problem are symmetrical, and the difference between two multiple solutions in terms of heat transfer may reach 34% for a given set of the governing parameters

Heating-induced inner-sphere substitution and reduction-oxidation reactions of the solid phenanthroline containing cobalt (2) and cobalt (3) complexes

International Nuclear Information System (INIS)

Palade, D.M.

1996-01-01

The results of the differential thermal and thermogravimetric analyses of solid phenanthroline-containing complexes of cobalt (2) and cobalt (3) in the atmosphere of the air have been analyzed. Mechanism of redox reactions occurring when cobalt (3) complexes are heated has been discussed. It is shown that some of gaseous products of the redox processes appear as a result of secondary reactions and not the processes of the ligands oxidation by Co 3+ . The influence of certain inner-sphere and coordinated anions (of I, inclusively) on cobalt (3) complexes behaviour during heating has been considered

Nuclear reactor containing facility

International Nuclear Information System (INIS)

Hidaka, Masataka; Murase, Michio.

1994-01-01

In a reactor containing facility, a condensation means is disposed above the water level of a cooling water pool to condensate steams of the cooling water pool, and return the condensated water to the cooling water pool. Upon occurrence of a pipeline rupture accident, steams generated by after-heat of a reactor core are caused to flow into a bent tube, blown from the exit of the bent tube into a suppression pool and condensated in a suppression pool water, thereby suppressing the pressure in the reactor container. Cooling water in the cooling water pool is boiled by heat conduction due to the condensation of steams, then the steams are exhausted to the outside of the reactor container to remove the heat of the reactor container to the outside of the reactor. In addition, since cooling water is supplied to the cooling water pool quasi-permanently by gravity as a natural force, the reactor container can be cooled by the cooling water pool for a long period of time. Since the condensation means is constituted with a closed loop and interrupted from the outside, radioactive materials are never released to the outside. (N.H.)

Heat transfer issues in high-heat-load synchrotron x-ray beams

International Nuclear Information System (INIS)

Khounsary, A.M.; Mills, D.M.

1994-09-01

In this paper, a short description of the synchrotron radiation x-ray sources and the associated power loads is given, followed by a brief description of typical synchrotron components and their heat load. It is emphasized that the design goals for most of these components is to limit (a) temperature, (b) stresses, or (c) strains in the system. Each design calls for a different geometry, material selection, and cooling scheme. Cooling schemes that have been utilized so far are primarily single phase and include simple macrochannel cooling, microchannel cooling, contact cooling, pin-post cooling, porous-flow cooling, jet cooling, etc. Water, liquid metals, and various cryogenic coolants have been used. Because the trend in x-ray beam development is towards brighter (i.e., more powerful) beams and assuming that no radical changes in the design of x-ray generating machines occurs in the next few years, it is fair to state that the utilization of various effective cooling schemes and, in particular, two-phase flow (e.g., subcooled boiling) warrants further investigation. This, however, requires a thorough examination of stability and reliability of two-phase flows for high-heat-flux components operating in ultrahigh vacuum with stringent reliability requirements

Steam condensation behavior of high pressure water's blow down directly into water in containment under LOCA

International Nuclear Information System (INIS)

Kusunoki, Tsuyoshi; Ishida, Toshihisa; Yoritsune, Tsutomu; Kasahara, Y.

1995-01-01

JAERI has been conducting a design study of an advanced type Marine Reactor X (MRX) for merchant ships. By employing 'Integral type PWR', In-vessel type control rod drive systems', 'Water filled containment system' and 'Decay heat removal system by natural convection', MRX achieved a compact, light weight and highly safe plant. Experiments on steam condensation behavior of high pressure water's blow down into water have been conducted in order to investigate a major safety issue related to the design decision of 'Water filled containment system'. (author)

Radioisotopic heat source

Science.gov (United States)

Jones, G.J.; Selle, J.E.; Teaney, P.E.

1975-09-30

Disclosed is a radioisotopic heat source and method for a long life electrical generator. The source includes plutonium dioxide shards and yttrium or hafnium in a container of tantalum-tungsten-hafnium alloy, all being in a nickel alloy outer container, and subjected to heat treatment of from about 1570$sup 0$F to about 1720$sup 0$F for about one h. (auth)

Device and Container for Reheating and Sterilization

Science.gov (United States)

Sastry, Sudhir K.; Heskitt, Brian F.; Jun, Soojin; Marcy, Joseph E.; Mahna, Ritesh

2012-01-01

Long-duration space missions require the development of improved foods and novel packages that do not represent a significant disposal issue. In addition, it would also be desirable if rapid heating technologies could be used on Earth as well, to improve food quality during a sterilization process. For this purpose, a package equipped with electrodes was developed that will enable rapid reheating of contents via ohmic heating to serving temperature during space vehicle transit. Further, the package is designed with a resealing feature, which enables the package, once used, to contain and sterilize waste, including human waste for storage prior to jettison during a long-duration mission. Ohmic heating is a technology that has been investigated on and off for over a century. Literature indicates that foods processed by ohmic heating are of superior quality to their conventionally processed counterparts. This is due to the speed and uniformity of ohmic heating, which minimizes exposure of sensitive materials to high temperatures. In principle, the material may be heated rapidly to sterilization conditions, cooled rapidly, and stored. The ohmic heating device herein is incorporated within a package. While this by itself is not novel, a reusable feature also was developed with the intent that waste may be stored and re-sterilized within the packages. These would then serve a useful function after their use in food processing and storage. The enclosure should be designed to minimize mass (and for NASA's purposes, Equivalent System Mass, or ESM), while enabling the sterilization function. It should also be electrically insulating. For this reason, Ultem high-strength, machinable electrical insulator was used.

Heat transfer modelling in the vertical tubes of a natural circulation passive containment loop with noncondensable gas

International Nuclear Information System (INIS)

Herranz, L.E.; Munoz-Cobo, J.L.; Tachenko, I.; Sancho, J.; Escriva, A.; Verdu, G.

1994-01-01

One of the key safety systems of the Simplified Boiling Water Reactor (SBWR) of General Electric is the Passive Containment Cooling System (PCCS). This system is designed to behave as a heat sink without need of operator actions in case of a reactor accident. Such a function relies on setting up a natural circulation loop between drywell and wetwell. Along this loop heat is removed by condensing the steam coming from the drywell onto the inner surface of externally cooled vertical tubes. Therefore, a successful design of the condenser requires a good knowledge of the local heat transmission coefficients. In this paper a model of steam condensation into vertical tubes is presented. Based on a modified diffusion boundary layer approach for noncondensables, this model accounts for the effect of shear stress caused by the cocurrent steam-gas mixture on the liquid film thickness. An approximate method to calculate film thickness, avoiding iterative algorithms, has been proposed. At present, this model has been implemented in HTCPIPE code and its results are being checked in terms of local heat transfer coefficients against the experimental data available. A good agreement between measurements and predictions is being observed for tests at atmospheric pressure. Further development and validation of the model is needed to consider aspects such as mist formation, wavy flow and high pressure. (author)

Roadmapping the Resolution of Gas Generation Issues in Packages Containing Radioactive Waste/Materials - A Status Report

International Nuclear Information System (INIS)

Luke, D.E.; Hamp, S.

2002-01-01

Gas generation issues, particularly hydrogen, have been an area of concern for the transport and storage of radioactive materials and waste in the Department of Energy (DOE) Complex. Potentially combustible gases can be generated through a variety of reactions, including chemical reactions and radiolytic decomposition of hydrogen-containing material. Since transportation regulations prohibit shipment of explosives and radioactive materials together, it was decided that hydrogen generation was a problem that warranted the execution of a high-level roadmapping effort. This paper discusses the major gas generation issues within the DOE Complex and the research that has been and is being conducted by the transuranic (TRU) waste, nuclear materials, and spent nuclear fuels (SNF) programs within DOE's Environmental Management (EM) organizations to address gas generation concerns. This paper presents a ''program level'' roadmap that links technology development to program needs and identifies the probability of success in an effort to understand the programmatic risk associated with the issue of gas generation. This paper also presents the status of the roadmap and follow-up activities

Heat transfer: Pittsburgh 1987

International Nuclear Information System (INIS)

Lyczkowski, R.W.

1987-01-01

This book contains papers divided among the following sections: Process Heat Transfer; Thermal Hydraulics and Phase Change Phenomena; Analysis of Multicomponent Multiphase Flow and Heat Transfer; Heat Transfer in Advanced Reactors; General Heat Transfer in Solar Energy; Numerical Simulation of Multiphase Flow and Heat Transfer; High Temperature Heat Transfer; Heat Transfer Aspects of Severe Reactor Accidents; Hazardous Waste On-Site Disposal; and General Papers

A container for containing and protecting a radioactive substance

International Nuclear Information System (INIS)

1974-01-01

The invention relates to a container adapted to contain and protect a radio-active substance. That container comprises a heat sensitive device for automatically (and, preferably, sealingly) enclosing and protecting the radio-active substance, should room temperature reach a predetermined level. Thus, the radio-active substance cannot escape in case of fire. Preferably, a bolt is also provided, capable of being actuated at a temperature slightly above the temperature actuating the protective device so as to maintain the radioactive substance protected. This can be applied to containers containing a radio-active substance such as polonium 210 [fr

Reactor containing facility

International Nuclear Information System (INIS)

Akagawa, Katsuhiko.

1992-01-01

A cooling space having a predetermined capacity is formed between a reactor container and concrete walls. A circulation loop disposed to the outside of the concrete walls is connected to the top and the bottom of the cooling space. The circulation loop has a circulation pump and a heat exchanger, and a cooling water supply pipe is connected to the upstream of the circulation pump for introducing cooling water from the outside. Upon occurrence of loss of coolant accident, cooling water is introduced from the cooling water supply pipe to the cooling space between the reactor container and the concrete walls after shut-down of the reactor operation. Then, cooling water is circulated while being cooled by the heat exchanger, to cool the reactor container by cooling water flown in the cooling space. This can cool the reactor container in a short period of time upon occurrence of the loss of coolant accident. Accordingly, a repairing operation for a ruptured portion can be conducted rapidly. (I.N.)

The log mean heat transfer rate method of heat exchanger considering the influence of heat radiation

International Nuclear Information System (INIS)

Wong, K.-L.; Ke, M.-T.; Ku, S.-S.

2009-01-01

The log mean temperature difference (LMTD) method is conventionally used to calculate the total heat transfer rate of heat exchangers. Because the heat radiation equation contains the 4th order exponential of temperature which is very complicate in calculations, thus LMTD method neglects the influence of heat radiation. From the recent investigation of a circular duct in some practical situations, it is found that even in the situation of the temperature difference between outer duct surface and surrounding is low to 1 deg. C, the heat radiation effect can not be ignored in the situations of lower ambient convective heat coefficient and greater surface emissivities. In this investigation, the log mean heat transfer rate (LMHTR) method which considering the influence of heat radiation, is developed to calculate the total heat transfer rate of heat exchangers.

Container for hydrogen isotopes

Energy Technology Data Exchange (ETDEWEB)

1976-01-12

A container is described for storage, shipping and and dispensing of hydrogen isotopes such as hydrogen, deuterium, tritium, or mixtures of the same. The container is compact, safe against fracture or accident, and is reusable. It consists of an outer housing with suitable inlet and outlet openings and electrical feed elements, the housing containing an activated sorber material in the form, for example of titanium sponge or an activated zirconium aluminate cartridge. The gas to be stored is introduced into the chamber under conditions of heat and vacuum and is retained in the sorber material. Subsequently, it may be released by heating the unit to drive off the stored gas at desired rates.

Container for hydrogen isotopes

International Nuclear Information System (INIS)

1976-01-01

A container is described for storage, shipping and and dispensing of hydrogen isotopes such as hydrogen, deuterium, tritium, or mixtures of the same. The container is compact, safe against fracture or accident, and is reusable. It consists of an outer housing with suitable inlet and outlet openings and electrical feed elements, the housing containing an activated sorber material in the form, for example of titanium sponge or an activated zirconium aluminate cartridge. The gas to be stored is introduced into the chamber under conditions of heat and vacuum and is retained in the sorber material. Subsequently, it may be released by heating the unit to drive off the stored gas at desired rates

Container for hydrogen isotopes

International Nuclear Information System (INIS)

Solomon, D.E.

1977-01-01

A container for the storage, shipping and dispensing of hydrogen isotopes such as hydrogen, deuterium, tritium, or mixtures of the same which has compactness, which is safe against fracture or accident, and which is reusable is described. The container consists of an outer housing with suitable inlet and outlet openings and electrical feed elements, the housing containing an activated sorber material in the form, for example, of titanium sponge or an activated zirconium aluminate cartridge. The gas to be stored is introduced into the chamber under conditions of heat and vacuum and will be retained in the sorber material. Subsequently, it may be released by heating the unit to drive off the stored gas at desired rates

An Overview of Liquid Fluoride Salt Heat Transport Systems

Energy Technology Data Exchange (ETDEWEB)

Holcomb, David Eugene [ORNL; Cetiner, Sacit M [ORNL

2010-09-01

, and this report also describes more recently developed technologies such as dry gas seals. This report also provides a high-level, parametric evaluation of LSHT loop performance to allow general intercomparisons between heat-transport fluid options as well as provide an overview of the properties and requirements for a representative loop. A compilation of relevant thermophysical properties of useful fluoride salts is also included for salt heat transport systems. Fluoride salts can be highly corrosive depending on the container materials selected, the salt chemistry, and the operating procedures used. The report includes an overview of the state-of-the-art in reduction-oxidation chemistry control methodologies employed to minimize corrosion issues. Salt chemistry control technology, however, remains at too low a level of understanding for widespread industrial usage. Loop operational issues such as start-up procedures and system freeze-up vulnerability are also discussed. Liquid fluoride salts are a leading candidate heat transport medium for high-temperature applications. This report provides an overview of the current status of liquid salt heat transport technology. The report includes a high-level, parametric evaluation of liquid fluoride salt heat transport loop performance to allow intercomparisons between heat-transport fluid options as well as providing an overview of the properties and requirements for a representative loop. Much of the information presented here derives from the earlier molten salt reactor program and a significant advantage of fluoride salts, as high temperature heat transport media is their consequent relative technological maturity. The report also includes a compilation of relevant thermophysical properties of useful heat transport fluoride salts. Fluoride salts are both thermally stable and with proper chemistry control can be relatively chemically inert. Fluoride salts can, however, be highly corrosive depending on the container

Heat pump planning handbook

CERN Document Server

Bonin, Jürgen

2015-01-01

The Heat Pump Planning Handbook contains practical information and guidance on the design, planning and selection of heat pump systems, allowing engineers, designers, architects and construction specialists to compare a number of different systems and options. Including detailed descriptions of components and their functions and reflecting the current state of technology this guide contains sample tasks and solutions as well as new model calculations and planning evaluations. Also economic factors and alternative energy sources are covered, which are essential at a time of rising heat costs. T

The new heating. Environmental-friendly and effectively heating with natural gas, wood, electricity and solar power; Die neue Heizung. Umweltfreundlich und wirtschaftlich heizen mit Gas, Holz, Strom und Sonnenenergie

Energy Technology Data Exchange (ETDEWEB)

Schulz, Marion; Westkaemper, Hubert

2013-02-01

Comfortable, climate-friendly and cost-effective heating technologies inclusive preparation of hot water and ventilation systems: Which technology is the best, most economical and most environmental one? Which fuel is economic as well as assured of a good future, if a new building or an energetically retrofitted old building is equipped with a heating system? These are the key issues for the house owners / building owners, planners, consultants and craftsmen. The book under consideration tries to answer these questions. Construction, planning benchmark figures and operational areas of modern heating technologies are described. The costs, the environmental pollutions and the economic efficiency of the most important systems are compared. The book contains hints on the decisive details of modern, energy saving system engineering including hot water preparation and ventilation technology. This handbook is suited for the selection and planning of energy saving and future-oriented heating systems.

Safety-analysis report for packaging (SARP) general-purpose heat-source module 750-Watt shipping container

International Nuclear Information System (INIS)

Whitney, M.A.; Burgan, C.E.; Blauvelt, R.K.; Zocher, R.W.; Bronisz, S.E.

1981-01-01

The SARP includes discussions of structural integrity, thermal resistance, radiation shielding and radiological safety, nuclear criticality safety, and quality control. Extensive tests and evaluations were performed to show that the container will function effectively with respect to all required standards and when subjected to normal transportation conditions and the sequence of four hypothetical accident conditions (free drop, puncture, thermal, and water immersion). In addition, a steady state temperature profile and radiation profile were measured using two heat sources that very closely resemble the GPHS. This gave an excellent representation of the GPHS temperature and radiation profile. A nuclear criticality safety analysis determined that all safety requirements are met

Experimental issues in in-situ synchrotron x-ray diffraction at high pressure and temperature by using a laser-heated diamond-anvil cell

International Nuclear Information System (INIS)

Yoo, C.S.

1997-01-01

An integrated technique of diamond-anvil cell, laser-heating and synchrotron x-ray diffraction technologies is capable of structural investigation of condensed matter in an extended region of high pressures and temperatures above 100 GPa and 3000 K. The feasibility of this technique to obtain reliable data, however, strongly depends on several experimental issues, including optical and x-ray setups, thermal gradients, pressure homogeneity, preferred orientation, and chemical reaction. In this paper, we discuss about these experimental issues together with future perspectives of this technique for obtaining accurate data

Integral Reactor Containment Condensation Model and Experimental Validation

Energy Technology Data Exchange (ETDEWEB)

Wu, Qiao [Oregon State Univ., Corvallis, OR (United States); Corradini, Michael [Univ. of Wisconsin, Madison, WI (United States)

2016-05-02

This NEUP funded project, NEUP 12-3630, is for experimental, numerical and analytical studies on high-pressure steam condensation phenomena in a steel containment vessel connected to a water cooling tank, carried out at Oregon State University (OrSU) and the University of Wisconsin at Madison (UW-Madison). In the three years of investigation duration, following the original proposal, the planned tasks have been completed: (1) Performed a scaling study for the full pressure test facility applicable to the reference design for the condensation heat transfer process during design basis accidents (DBAs), modified the existing test facility to route the steady-state secondary steam flow into the high pressure containment for controllable condensation tests, and extended the operations at negative gage pressure conditions (OrSU). (2) Conducted a series of DBA and quasi-steady experiments using the full pressure test facility to provide a reliable high pressure condensation database (OrSU). (3) Analyzed experimental data and evaluated condensation model for the experimental conditions, and predicted the prototypic containment performance under accidental conditions (UW-Madison). A film flow model was developed for the scaling analysis, and the results suggest that the 1/3 scaled test facility covers large portion of laminar film flow, leading to a lower average heat transfer coefficient comparing to the prototypic value. Although it is conservative in reactor safety analysis, the significant reduction of heat transfer coefficient (50%) could under estimate the prototypic condensation heat transfer rate, resulting in inaccurate prediction of the decay heat removal capability. Further investigation is thus needed to quantify the scaling distortion for safety analysis code validation. Experimental investigations were performed in the existing MASLWR test facility at OrST with minor modifications. A total of 13 containment condensation tests were conducted for pressure

Design of megawatt power level heat pipe reactors

Energy Technology Data Exchange (ETDEWEB)

Mcclure, Patrick Ray [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Poston, David Irvin [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Dasari, Venkateswara Rao [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Reid, Robert Stowers [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

2015-11-12

An important niche for nuclear energy is the need for power at remote locations removed from a reliable electrical grid. Nuclear energy has potential applications at strategic defense locations, theaters of battle, remote communities, and emergency locations. With proper safeguards, a 1 to 10-MWe (megawatt electric) mobile reactor system could provide robust, self-contained, and long-term power in any environment. Heat pipe-cooled fast-spectrum nuclear reactors have been identified as a candidate for these applications. Heat pipe reactors, using alkali metal heat pipes, are perfectly suited for mobile applications because their nature is inherently simpler, smaller, and more reliable than “traditional” reactors. The goal of this project was to develop a scalable conceptual design for a compact reactor and to identify scaling issues for compact heat pipe cooled reactors in general. Toward this goal two detailed concepts were developed, the first concept with more conventional materials and a power of about 2 MWe and a the second concept with less conventional materials and a power level of about 5 MWe. A series of more qualitative advanced designs were developed (with less detail) that show power levels can be pushed to approximately 30 MWe.

Estimation of time to rupture in a fire using 6FIRE, a lumped parameter UF6 cylinder transient heat transfer/stress analysis model

International Nuclear Information System (INIS)

Williams, W.R.; Anderson, J.C.

1995-01-01

The transportation of UF 6 is subject to regulations requiring the evaluation of packaging under a sequence of hypothetical accident conditions including exposure to a 30-min 800 degree C (1475 degree F) fire [10 CFR 71.73(c)(3)]. An issue of continuing interest is whether bare cylinders can withstand such a fire without rupturing. To address this issue, a lumped parameter heat transfer/stress analysis model (6FIRE) has been developed to simulate heating to the point of rupture of a cylinder containing UF 6 when it is exposed to a fire. The model is described, then estimates of time to rupture are presented for various cylinder types, fire temperatures, and fill conditions. An assessment of the quantity of UF 6 released from containment after rupture is also presented. Further documentation of the model is referenced

US-Japan workshop Q-181 on high heat flux components and plasma-surface interactions for next devices: Proceedings

Energy Technology Data Exchange (ETDEWEB)

McGrath, R.T. [ed.] [Sandia National Labs., Albuquerque, NM (United States); Yamashina, T. [ed.] [Hokkadio Univ. (Japan)

1994-04-01

This report contain viewgraphs of papers from the following sessions: plasma facing components issues for future machines; recent PMI results from several tokamaks; high heat flux technology; plasma facing components design and applications; plasma facing component materials and irradiation damage; boundary layer plasma; plasma disruptions; conditioning and tritium; and erosion/redeposition.

US-Japan workshop Q-181 on high heat flux components and plasma-surface interactions for next devices: Proceedings

International Nuclear Information System (INIS)

McGrath, R.T.; Yamashina, T.

1994-04-01

This report contain viewgraphs of papers from the following sessions: plasma facing components issues for future machines; recent PMI results from several tokamaks; high heat flux technology; plasma facing components design and applications; plasma facing component materials and irradiation damage; boundary layer plasma; plasma disruptions; conditioning and tritium; and erosion/redeposition

Feasibility study of cocos, condensation of containment atmosphere on structures

International Nuclear Information System (INIS)

Rij, H.M. van; Vonka, V.

1989-12-01

The aim of this report is to assess the state of the art of the knowledge of the thermo-hydraulic conditions within a LWR containment in order to determine both the radioactive and the non-radioactive aerosol deposition rates during a severe reactor accident. The radioactive aerosol in the containment atmosphere is, together with the noble gases, responsible for the radioactive source term into the biosphere when a containment failure occurs. The dominant aerosol removal mechanisms depend strongly upon the thermal-hydraulic state of the containment atmosphere. It is demonstrated that the thermohydraulic state, determined by heat sources and the sensible heat transport, is predominantly super-heated when fission products are released into the containment. Hence the thermohydraulic conditions are not favorable for an intensive bulk condensation onto aerosol particles during an essential period of time. A station black-out scenario, in which the primary system of the considered 500 MWe PWR with a dry cavity is depressurized prior to vessel failure, is used as an example to demonstrate this effect. The results, obtained with the CONTAIN code, show the relevance of the sensible heat transport in the period of time (c.a. 30 minutes) between the end of the injection of the steam and fission products into the containment, and the molten core concrete interaction. All considered variation of the station black-out scenario, in which the decay heat dissipated in the containment atmosphere has been 10% of the total decay heat, indicate that the atmosphere becomes super-heated within the 30 minutes. Reducing the fraction of the decay heat in the containment from the 10% to 5% increases the time period with saturated conditions. The amount of the decay heat released into the containment atmosphere forms the major factor determining the thermohydraulic state. It influences the duration of the bulk condensation period, which in turn has an influence on the aerosol deposition

Drug Abuse on College Campuses: Emerging Issues. Issues in Prevention

Science.gov (United States)

Higher Education Center for Alcohol, Drug Abuse, and Violence Prevention, 2012

2012-01-01

This "Issues in Prevention" focuses on emerging issues concerning drug abuse on college campuses. This issue contains the following articles: (1) Drug Abuse Trends; (2) Q&A With Jim Lange; (3) Bath Salts; (4) Refuse to Abuse; (5) Related Federal Resource; and (6) Higher Education Center Resources.

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

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.)

NPR and ANSI Containment Study Using Passive Cooling Techniques

International Nuclear Information System (INIS)

Shin, J. J.; Iotti, R. C.; Wright, R. F.

1993-01-01

Passive containment cooling study of NPR (New Production Reactor) and ANSI (Advanced Neutron Source) following postulated loss of coolant accident with a coincident station blackout due to total loss of all alternating current power are studied analytically and experimentally. All the reactor and containment cooling under this condition would rely on the passive cooling system which removes reactor decay heat and provides emergency core and containment cooling. Containment passive emergency core and containment cooling. Containment passive cooling for this study takes place in the annulus between containment steel shell and concrete shield building by natural convection air flow and concrete shield building by natural convection air flow and thermal radiation. Various heat transfer coefficients inside annular air space were investigated by running the modified Contempt code Contempt-Npr. In order to verify proper heat transfer coefficient, temperature, heat flux and velocity profiles were measured inside annular air space of the test facility which is a 24 foot (7.3m) high, steam heated inner cylinder of three foot (.91m) diameter and five and halt foot (1.7m) diameter outer cylinder. Comparison of Contempt-Npr and WGOTHIC was done for reduced scale Npr. It is concluded that Npr and ANSI containments can be passively cooled with air alone without extended cooling surfaces or passive water spray

Nuclear power plant with a containment

International Nuclear Information System (INIS)

Barthelmes, C.P.

1982-01-01

In nuclear power plants there is usually a containment incorporating components bearing activity. If in the cladding free hydrogen develops, controlled oxidation must be ensured by means of a recombination device, in order to prevent oxyhydrogen explosions. For this purpose, a permanent thoroughmixing of the gases in the containment is required. This can be achieved by vertical shafts reaching to at least half the height of the containment and provided with heating devices to initiate the gas circulation by the stack effect. These heating devices mainly serve as thermal recombinator. (orig.) [de

Evaporation of liquefied natural gas in conditions of compact storage containers heating

Science.gov (United States)

Telgozhayeva, D. S.

2014-08-01

Identical by its power, but located in different parts of the external surface of the tank, the heating sources are different intensity heat transfer modes is heating up, respectively, times of vapour pressure rise to critical values. Developed mathematical model and method of calculation can be used in the analysis of conditions of storage tanks for liquefied gases.

Kinetic Hydration Heat Modeling for High-Performance Concrete Containing Limestone Powder

Directory of Open Access Journals (Sweden)

Xiao-Yong Wang

2017-01-01

Full Text Available Limestone powder is increasingly used in producing high-performance concrete in the modern concrete industry. Limestone powder blended concrete has many advantages, such as increasing the early-age strength, reducing the setting time, improving the workability, and reducing the heat of hydration. This study presents a kinetic model for modeling the hydration heat of limestone blended concrete. First, an improved hydration model is proposed which considers the dilution effect and nucleation effect due to limestone powder addition. A degree of hydration is calculated using this improved hydration model. Second, hydration heat is calculated using the degree of hydration. The effects of water to binder ratio and limestone replacement ratio on hydration heat are clarified. Third, the temperature history and temperature distribution of hardening limestone blended concrete are calculated by combining hydration model with finite element method. The analysis results generally agree with experimental results of high-performance concrete with various mixing proportions.

The German Final Repository Konrad for Low and Intermediate Level Waste with Negligible Heat Generation - Water Law Issues

International Nuclear Information System (INIS)

Boetsch, W.; Grundler, D.; Kugel, K.; Brennecke, P.; Steyer, S.

2009-01-01

A survey on the conceptual realization of the requirements due to water law aspects within the license the KONRAD repository for radioactive waste with negligible heat generation in Germany is given [1]. The regulatory decision for the implementation and operation of the repository KONRAD includes, among other things, water law issues. In particular, the KONRAD license includes waste requirements concerning non-radioactive hazardous material (waste package constituents) which have to be considered producing KONRAD waste packages. The intended philosophy of waste acceptance and waste package quality assurance measures to be considered by the KONRAD site operator as well as by the waste producer will be presented. It will demonstrate the selected procedure of the waste declaration and acceptance and describe the structure and logic of tools and aids to comply with the legal requirements of the license and its collateral clause issued under water law. (authors)

Local Heat Transfer and CHF for Subcooled Flow Boiling - Annual Report 1993

International Nuclear Information System (INIS)

Boyd, Ronald D.

2000-01-01

Subcooled flow boiling in heated coolant channels is an important heat transfer enhancement technique in the development of fusion reactor components, where high heat fluxes must be accommodated. As energy fluxes increase in magnitude, additional emphasis must be devoted to enhancing techniques such as sub cooling and enhanced surfaces. In addition to subcooling, other high heat flux alternatives such as high velocity helium and liquid metal cooling have been considered as serious contenders. Each technique has its advantages and disadvantages [1], which must be weighed as to reliability and reduced cost of fusion reactor components. Previous studies [2] have set the stage for the present work, which will concentrate on fundamental thermal hydraulic issues associated with the h-international Thermonuclear Experimental Reactor (ITER) and the Engineering Design Activity (EDA). This proposed work is intended to increase our understanding of high heat flux removal alternatives as well as our present capabilities by: (1) including single-side heating effects in models for local predictions of heat transfer and critical heat flux; (2) inspection of the US, Japanese, and other possible data sources for single-side heating, with the aim of exploring possible correlations for both CHF and local heat transfer; and (3) assessing the viability of various high heat flux removal techniques. The latter task includes: (a) sub-cooled water flow boiling with enhancements such as twisted tapes, and hypervapotrons, (b) high velocity helium cooling, and (c) other potential techniques such as liquid metal cooling. This assessment will increase our understanding of: (1) hypervapotron heat transfer via fins, flow recirculation, and flow oscillation, and (2) swirl flow. This progress report contains selective examples of ongoing work. Section II contains an extended abstract, which is part of and evolving technical paper on single-side f heating. Section III describes additional details

A Conceptual Change Model for Teaching Heat Energy, Heat Transfer and Insulation

Science.gov (United States)

Lee, C. K.

2014-01-01

This study examines the existing knowledge that pre-service elementary teachers (PSETs) have regarding heat energy, heat transfer and insulation. The PSETs' knowledge of heat energy was initially assessed by using an activity: determining which container would be best to keep hot water warm for the longest period of time. Results showed that PSETs…

Transient shielded liquid hydrogen containers

International Nuclear Information System (INIS)

Varghese, A.P.; Herring, R.H.

1990-01-01

The storage of hydrogen in the liquid phase has been limited in duration due to the thermal performance constraints of conventional Liquid Hydrogen containers available. Conventional Liquid Hydrogen containers lose hydrogen because of their relatively high heat leak and variations in usage pattern of hydrogen due to shutdowns. Local regulations also discourage venting of hydrogen. Long term storage of Liquid Hydrogen without product loss was usually accomplished using Liquid Nitrogen sacrificial shields. This paper reports on a new low heat leak container developed and patented that will extend the storage time of liquid hydrogen by five hundred percent. The principle of operation of the Transient Shields which makes the extraordinary performance of this container feasible is described in this paper. Also covered are the impact of this new container on present applications of hydrogen and the new opportunities afforded to Liquid hydrogen in the world hydrogen market

Pressure suppression facility for reactor container

International Nuclear Information System (INIS)

Fujii, Tadashi; Fukui, Toru; Kataoka, Yoshiyuki; Tominaga, Kenji.

1993-01-01

In a nuclear reactor comprising heat transfer surfaces from a pressure suppression pool at the inside to the outer circumferential pool at the outside, a means for supplying water from a water supply source at the outside of the container to the pools is disposed. Then, a heat transfer means is disposed between the pressure suppression chamber and the water cooling pool. The water supply means comprises a pressurization means for applying pressure to water of the water supply source and a water supply channel. Water is supplied into the pressure suppression pool and the outer circumferential pool to elevate the water level and extend the region of heat contact with the water cooling heat transfer means. In addition, since dynamic pressure is applied to the feedwater, for example, by pressurizing the water surface of the water supply source, water can be supplied without using dynamic equipments such as pumps. Then, since water-cooling heat transfer surface can be extended after occurrence of accident, enlargement of a reactor container and worsening of earthquake proofness can be avoided as much as possible, to improve function for suppressing the pressure in the container. Further, since water-cooling heat transfer region can be extended, the arrangement of the water source and the place to which water is supplied is made optional without considering the relative height therebetween, to improve earthquake proofness. (N.H.)

77 FR 33486 - Certain Integrated Circuit Packages Provided With Multiple Heat-Conducting Paths and Products...

Science.gov (United States)

2012-06-06

...Notice is hereby given that the U.S. International Trade Commission has received a complaint entitled Certain Integrated Circuit Packages Provided With Multiple Heat-Conducting Paths and Products Containing Same, DN 2899; the Commission is soliciting comments on any public interest issues raised by the complaint or complainant's filing under section 210.8(b) of the Commission's Rules of Practice and Procedure (19 CFR 210.8(b)).

Regenerative adsorbent heat pump

Science.gov (United States)

Jones, Jack A. (Inventor)

1991-01-01

A regenerative adsorbent heat pump process and system is provided which can regenerate a high percentage of the sensible heat of the system and at least a portion of the heat of adsorption. A series of at least four compressors containing an adsorbent is provided. A large amount of heat is transferred from compressor to compressor so that heat is regenerated. The process and system are useful for air conditioning rooms, providing room heat in the winter or for hot water heating throughout the year, and, in general, for pumping heat from a lower temperature to a higher temperature.

District heating in Italy

International Nuclear Information System (INIS)

Sacchi, E.

1998-01-01

The legislative act establishing the electric monopoly virtually shut out the district heating associated with electricity cogeneration, while other laws, issued to counteract the effects of oil shocks, allowed municipal utilities to do so. Thus, district heating has experienced some development, though well below its possibilities. The article analyses the reasons for this lagging, reports district heating data and projects its forecasts against the Kyoto Protocol objectives [it

Depressurization as an accident management strategy to minimize the consequences of direct containment heating

International Nuclear Information System (INIS)

Hanson, D.J.; Golden, D.W.; Chambers, R.; Miller, J.D.; Hallbert, B.P.; Dobbe, C.A.

1990-10-01

Probabilistic Risk Assessments (PRAs) have identified severe accidents for nuclear power plants that have the potential to cause failure of the containment through direct containment heating (DCH). Prevention of DCH or mitigation of its effects may be possible using accident management strategies that intentionally depressurize the reactor coolant system (RCS). The effectiveness of intentional depressurization during a station blackout TMLB' sequence was evaluated considering the phenomenological behavior, hardware performance, and operational performance. Phenomenological behavior was calculated using the SCDAP/RELAP5 severe accident analysis code. Two strategies to mitigate DCH by depressurization of the RCS were considered. One strategy, called early depressurization, assumed that the reactor head vent and pressurizer power-operated relief valves (PORVs) were latched open at steam generator dryout. The second strategy, called late depression, assumed that the head vent and PORVs were latched open at a core exit temperature of ∼922 K (1200 degree F). Depressurization of the RCS to a low value that may mitigate DCH was predicted prior to reactor pressure vessel breach for both early and late depressurization. The strategy of late depressurization is preferred over early depressurization because there are greater opportunities to recover plant functions prior to core damage and because failure uncertainties are lessened. 22 refs., 38 figs., 6 tabs

Depressurization as an accident management strategy to minimize the consequences of direct containment heating

Energy Technology Data Exchange (ETDEWEB)

Hanson, D.J.; Golden, D.W.; Chambers, R.; Miller, J.D.; Hallbert, B.P.; Dobbe, C.A. (EG and G Idaho, Inc., Idaho Falls, ID (USA))

1990-10-01

Probabilistic Risk Assessments (PRAs) have identified severe accidents for nuclear power plants that have the potential to cause failure of the containment through direct containment heating (DCH). Prevention of DCH or mitigation of its effects may be possible using accident management strategies that intentionally depressurize the reactor coolant system (RCS). The effectiveness of intentional depressurization during a station blackout TMLB' sequence was evaluated considering the phenomenological behavior, hardware performance, and operational performance. Phenomenological behavior was calculated using the SCDAP/RELAP5 severe accident analysis code. Two strategies to mitigate DCH by depressurization of the RCS were considered. One strategy, called early depressurization, assumed that the reactor head vent and pressurizer power-operated relief valves (PORVs) were latched open at steam generator dryout. The second strategy, called late depression, assumed that the head vent and PORVs were latched open at a core exit temperature of {approximately}922 K (1200{degree}F). Depressurization of the RCS to a low value that may mitigate DCH was predicted prior to reactor pressure vessel breach for both early and late depressurization. The strategy of late depressurization is preferred over early depressurization because there are greater opportunities to recover plant functions prior to core damage and because failure uncertainties are lessened. 22 refs., 38 figs., 6 tabs.

Corrosion-Resistant Container for Molten-Material Processing

Science.gov (United States)

Stern, Theodore G.; McNaul, Eric

2010-01-01

In a carbothermal process, gaseous methane is passed over molten regolith, which is heated past its melting point to a temperature in excess of 1,625 C. At this temperature, materials in contact with the molten regolith (or regolith simulant) corrode and lose their structural properties. As a result, fabricating a crucible to hold the molten material and providing a method of contact heating have been problematic. Alternative containment approaches use a large crucible and limit the heat zone of the material being processed, which is inefficient because of volume and mass constraints. Alternative heating approaches use non-contact heating, such as by laser or concentrated solar energy, which can be inefficient in transferring heat and thus require higher power heat sources to accomplish processing. The innovation is a combination of materials, with a substrate material having high structural strength and stiffness and high-temperature capability, and a coating material with a high corrosion resistance and high-temperature capability. The material developed is a molybdenum substrate with an iridium coating. Creating the containment crucible or heater jacket using this material combination requires only that the molybdenum, which is easily processed by conventional methods such as milling, electric discharge machining, or forming and brazing, be fabricated into an appropriate shape, and that the iridium coating be applied to any surfaces that may come in contact with the corrosive molten material. In one engineering application, the molybdenum was fashioned into a container for a heat pipe. Since only the end of the heat pipe is used to heat the regolith, the container has a narrowing end with a nipple in which the heat pipe is snugly fit, and the external area of this nipple, which contacts the regolith to transfer heat into it, is coated with iridium. At the time of this reporting, no single material has been found that can perform the functions of this combination

Integrated analysis of core debris interactions and their effects on containment integrity using the CONTAIN computer code

International Nuclear Information System (INIS)

Carroll, D.E.; Bergeron, K.D.; Williams, D.C.; Tills, J.L.; Valdez, G.D.

1987-01-01

The CONTAIN computer code includes a versatile system of phenomenological models for analyzing the physical, chemical and radiological conditions inside the containment building during severe reactor accidents. Important contributors to these conditions are the interactions which may occur between released corium and cavity concrete. The phenomena associated with interactions between ejected corium debris and the containment atmosphere (Direct Containment Heating or DCH) also pose a potential threat to containment integrity. In this paper, we describe recent enhancements of the CONTAIN code which allow an integrated analysis of these effects in the presence of other mitigating or aggravating physical processes. In particular, the recent inclusion of the CORCON and VANESA models is described and a calculation example presented. With this capability CONTAIN can model core-concrete interactions occurring simultaneously in multiple compartments and can couple the aerosols thereby generated to the mechanistic description of all atmospheric aerosol components. Also discussed are some recent results of modeling the phenomena involved in Direct Containment Heating. (orig.)

Cooling device for reactor container

International Nuclear Information System (INIS)

Arai, Kenji.

1996-01-01

Upon assembling a static container cooling system to an emergency reactor core cooling system using dynamic pumps in a power plant, the present invention provides a cooling device of lowered center of gravity and having a good cooling effect by lowering the position of a cooling water pool of the static container cooling system. Namely, the emergency reactor core cooling system injects water to the inside of a pressure vessel using emergency cooling water stored in a suppression pool as at least one water source upon loss of reactor coolant accident. In addition, a cooling water pool incorporating a heat exchanger is disposed at the circumference of the suppression pool at the outside of the container. A dry well and the heat exchanger are connected by way of steam supply pipes, and the heat exchanger is connected with the suppression pool by way of a gas exhaustion pipe and a condensate returning pipeline. With such a constitution, the position of the heat exchanger is made higher than an ordinary water level of the suppression pool. As a result, the emergency cooling water of the suppression pool water is injected to the pressure vessel by the operation of the reactor cooling pumps upon loss of coolant accident to cool the reactor core. (I.S.)

Internally Heated Screw Pyrolysis Reactor (IHSPR) heat transfer performance study

Science.gov (United States)

Teo, S. H.; Gan, H. L.; Alias, A.; Gan, L. M.

2018-04-01

1.5 billion end-of-life tyres (ELT) were discarded globally each year and pyrolysis is considered the best solution to convert the ELT into valuable high energy-density products. Among all pyrolysis technologies, screw reactor is favourable. However, conventional screw reactor risks plugging issue due to its lacklustre heat transfer performance. An internally heated screw pyrolysis reactor (IHSPR) was developed by local renewable energy industry, which serves as the research subject for heat transfer performance study of this particular paper. Zero-load heating test (ZLHT) was first carried out to obtain the operational parameters of the reactor, followed by the one dimensional steady-state heat transfer analysis carried out using SolidWorks Flow Simulation 2016. Experiments with feed rate manipulations and pyrolysis products analyses were conducted last to conclude the study.

Net Weight Issue LLNL DOE-STD-3013 Containers

International Nuclear Information System (INIS)

Wilk, P

2008-01-01

The following position paper will describe DOE-STD-3013 container sets No.L000072 and No.L000076, and how they are compliant with DOE-STD-3013-2004. All masses of accountable nuclear materials are measured on LLNL certified balances maintained under an MC and A Program approved by DOE/NNSA LSO. All accountability balances are recalibrated annually and checked to be within calibration on each day that the balance is used for accountability purposes. A statistical analysis of the historical calibration checks from the last seven years indicates that the full-range Limit of Error (LoE, 95% confidence level) for the balance used to measure the mass of the contents of the above indicated 3013 containers is 0.185 g. If this error envelope, at the 95% confidence level, were to be used to generate an upper-limit to the measured weight of the containers No.L000072 and No.L000076, the error-envelope would extend beyond the 5.0 kg 3013-standard limit on the package contents by less than 0.3 g. However, this is still well within the intended safety bounds of DOE-STD-3013-2004

Heated Tube Facility

Data.gov (United States)

Federal Laboratory Consortium — The Heated Tube Facility at NASA GRC investigates cooling issues by simulating conditions characteristic of rocket engine thrust chambers and high speed airbreathing...

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

CFD simulations in the nuclear containment using the DES turbulence models

International Nuclear Information System (INIS)

Ding, Peng; Chen, Meilan; Li, Wanai; Liu, Yulan; Wang, Biao

2015-01-01

Highlights: • The k-ε based DES model is used in the nuclear containment simulation. • The comparison of results between different turbulent models is obtained. • The superiority of DES models is analyzed. • The computational efficiency with the DES turbulence models is explained. - Abstract: Different species of gases would be released into the containment and cause unpredicted disasters during the nuclear severe accidents. It is important to accurately predict the transportation and stratification phenomena of these gas mixtures. CFD simulations of these thermal hydraulic issues in nuclear containment are investigated in this paper. The main work is to study the influence of turbulence model on the calculation of gas transportation and heat transfer. The k-ε based DES and other frequently used turbulence models are used in the steam and helium release simulation in THAI series experiment. This paper will show the superiority of the DES turbulence model in terms of computational efficiency and accuracy with the experimental results, and analyze the necessities of DES model to simulate the large-scale containment flows with both laminar and turbulence regions

CFD simulations in the nuclear containment using the DES turbulence models

Energy Technology Data Exchange (ETDEWEB)

Ding, Peng [School of Engineering, Sun Yat-Sen University, Guangzhou (China); Chen, Meilan [China Nuclear Power Technology Research Institute, Shenzhen (China); Li, Wanai, E-mail: liwai@mail.sysu.edu.cn [Sino-French Institute of Nuclear Engineering & Technology, Sun Yat-Sen University, Guangzhou (China); Liu, Yulan [School of Engineering, Sun Yat-Sen University, Guangzhou (China); Wang, Biao [Sino-French Institute of Nuclear Engineering & Technology, Sun Yat-Sen University, Guangzhou (China)

2015-06-15

Highlights: • The k-ε based DES model is used in the nuclear containment simulation. • The comparison of results between different turbulent models is obtained. • The superiority of DES models is analyzed. • The computational efficiency with the DES turbulence models is explained. - Abstract: Different species of gases would be released into the containment and cause unpredicted disasters during the nuclear severe accidents. It is important to accurately predict the transportation and stratification phenomena of these gas mixtures. CFD simulations of these thermal hydraulic issues in nuclear containment are investigated in this paper. The main work is to study the influence of turbulence model on the calculation of gas transportation and heat transfer. The k-ε based DES and other frequently used turbulence models are used in the steam and helium release simulation in THAI series experiment. This paper will show the superiority of the DES turbulence model in terms of computational efficiency and accuracy with the experimental results, and analyze the necessities of DES model to simulate the large-scale containment flows with both laminar and turbulence regions.

Sanskrit Issue.

Science.gov (United States)

Miller, Barbara Stoler, Ed.

1971-01-01

This issue of "Mahfil" is devoted to Sanskrit literature and contains a note on Sanskrit pronunciation and selections of Sanskrit literature. It also contains articles analyzing and discussing various aspects of the literature, including "Sanskrit Rhetoric and Poetic,""The Creative Role of the Goddess Vac in the…

Plasma heating

International Nuclear Information System (INIS)

Wilhelm, R.

1989-01-01

Successful plasma heating is essential in present fusion experiments, for the demonstration of DpT burn in future devices and finally for the fusion reactor itself. This paper discusses the common heating systems with respect to their present performance and their applicability to future fusion devices. The comparative discussion is oriented to the various function of heating, which are: - plasma heating to fusion-relevant parameters and to ignition in future machines, -non-inductive, steady-pstate current drive, - plasma profile control, -neutral gas breakdown and plasma build-up. In view of these different functions, the potential of neutral beam injection (NBI) and the various schemes of wave heating (ECRH, LH, ICRH and Alven wave heating) is analyzed in more detail. The analysis includes assessments of the present physical and technical state of these heating methods, and makes suggestions for future developments and about outstanding problems. Specific attention is given to the still critical problem of efficient current drive, especially with respect to further extrapolation towards an economically operating tokamak reactor. Remarks on issues such as reliability, maintenance and economy conclude this comparative overview on plasma heating systems. (author). 43 refs.; 13 figs.; 3 tabs

Containment Modelling with the ASTEC Code

International Nuclear Information System (INIS)

Sadek, Sinisa; Grgic, Davor

2014-01-01

ASTEC is an integral computer code jointly developed by Institut de Radioprotection et de Surete Nucleaire (IRSN, France) and Gesellschaft fur Anlagen-und Reaktorsicherheit (GRS, Germany) to assess the nuclear power plant behaviour during a severe accident (SA). It consists of 13 coupled modules which compute various SA phenomena in primary and secondary circuits of the nuclear power plants (NPP), and in the containment. The ASTEC code was used to model and to simulate NPP behaviour during a postulated station blackout accident in the NPP Krsko, a two-loop pressurized water reactor (PWR) plant. The primary system of the plant was modelled with 110 thermal hydraulic (TH) volumes, 113 junctions and 128 heat structures. The secondary system was modelled with 76 TH volumes, 77 junctions and 87 heat structures. The containment was modelled with 10 TH volumes by taking into account containment representation as a set of distinctive compartments, connected with 23 junctions. A total of 79 heat structures were used to simulate outer containment walls and internal steel and concrete structures. Prior to the transient calculation, a steady state analysis was performed. In order to achieve correct plant initial conditions, the operation of regulation systems was modelled. Parameters which were subjected to regulation were the pressurizer pressure, the pressurizer narrow range level and steam mass flow rates in the steam lines. The accident analysis was focused on containment behaviour, however the complete integral NPP analysis was carried out in order to provide correct boundary conditions for the containment calculation. During the accident, the containment integrity was challenged by release of reactor system coolant through degraded coolant pump seals and, later in the accident following release of the corium out of the reactor pressure vessel, by the molten corium concrete interaction and direct containment heating mechanisms. Impact of those processes on relevant

Materials compatibility issues related to thermal energy storage for a space solar dynamic power system

Science.gov (United States)

Faget, N. M.

1986-01-01

Attention is given to results obtained to date in developmental investigations of a thermal energy storage (TES) system for the projected NASA Space Station's solar dynamic power system; these tests have concentrated on issues related to materials compatibility for phase change materials (PCMs) and their containment vessels' materials. The five PCMs tested have melting temperatures that correspond to the operating temperatures of either the Brayton or Rankine heat engines, which were independently chosen for their high energy densities.

Heat transfer analyses using computational fluid dynamics in the air blast freezing of guava pulp in large containers

Directory of Open Access Journals (Sweden)

W. M. Okita

2013-12-01

Full Text Available Heat transfer during the freezing of guava pulp conditioned in large containers such as in stacked boxes (34 L and buckets (20 L and unstacked drums (200 L is discussed. The air velocities across the cross-section of the tunnel were measured, and the values in the outlet of the evaporator were used as the initial conditions in computational fluid dynamics (CFD simulations. The model tested was turbulent standard k-ε. The CFD-generated convective heat transfer coefficients were mapped on the surfaces for each configuration and used in procedures for the calculation of freezing-time estimates. These estimates were compared with the experimental results for validation. The results showed that CFD determined representative coefficients and produced good correlations between the predicted and experimental values when applied to the freezing-time estimates for the box and drum configurations. The errors depended on the configuration and the adopted mesh (3-D grid construction.

Heating ability and biocompatibility study of silica-coated magnetic ...

Indian Academy of Sciences (India)

Home; Journals; Bulletin of Materials Science; Volume 38; Issue 6. Heating ability and biocompatibility study of silica-coated magnetic nanoparticles as heating mediators for magnetic hyperthermia and magnetically triggered drug delivery systems. Meysam Soleymani Mohammad Edrissi. Volume 38 Issue 6 October 2015 ...

Analysis and proposal of implementation effects of heat metering and energy efficiency retrofit of existing residential buildings in northern heating areas of China in “the 11th Five-Year Plan” period

International Nuclear Information System (INIS)

Bao Lingling; Zhao Jing; Zhu Neng

2012-01-01

In China, northern heating region contains approximately 6.5 billion m 2 residential building areas accounting for 15% of the total residential living areas of urban and rural. About 70% of the urban residential buildings in north China are high energy consumption buildings. The task of heat metering and energy efficiency retrofit of 0.15 billion m 2 existing residential buildings in northern heating areas of China in “the 11th Five-Year Plan” period was proposed by the Ministry of Housing and Urban–Rural Development (MOHURD) in 2007 and completed in 2010. This paper introduced both central and local governments' efforts on organization, implementation and finance, etc. Then several retrofitting effects involving improving the people's livelihood, mobilizing the enthusiasm of residents for the retrofit and driving the development of relevant industries were presented. Finally, on the basis of analyzing the issues encountered in the progress of the retrofit in the past 4 years, the paper gave some policy proposals on organization system, financing models, reward mechanism, and heating system reformation to help to promote the energy efficiency retrofit in “the 12th Five-Year Plan” period. - Highlights: ► Specific approaches of heat metering and energy efficiency retrofit (HMEER) at central and local level are introduced. ► Main HMEER effects are presented. ► Analyzing several issues encountered in the progress of the HMEER. ► Corresponding proposals are provided.

Evaluation of Station Blackout accidents at nuclear power plants. Technical findings related to Unresolved Safety Issue A-44. Draft report for comment

International Nuclear Information System (INIS)

Baranowsky, P.W.

1985-05-01

''Station Blackout,'' which is the complete loss of alternating current (ac) electrical power in a nuclear power plant, has been designated as Unresolved Safety Issue A-44. Because many safety systems required for reactor core decay heat removal and containment heat removal depend on ac power, the consequences of a station blackout could be severe. This report documents the findings of technical studies performed as part of the program to resolve this issue. The important factors analyzed include: the frequency of loss of offsite power; the probability that emergency or onsite ac power supplies would be unavailable; the capability and reliability of decay heat removal systems independent of ac power; and the likelihood that offsite power would be restored before systems that cannot operate for extended periods without ac power fail, thus resulting in core damage. This report also addresses effects of different designs, locations, and operational features on the estimated frequency of core damage resulting from station blackout events

Molten material-containing vessel

International Nuclear Information System (INIS)

Akagawa, Katsuhiko

1998-01-01

The molten material-containing vessel of the present invention comprises a vessel main body having an entrance opened at the upper end, a lid for closing the entrance, an outer tube having an upper end disposed at the lower surface of the lid, extended downwardly and having an closed lower end and an inner tube disposed coaxially with the outer tube. When a molten material is charged from the entrance to the inside of the vessel main body of the molten material-containing vessel and the entrance is closed by the lid, the outer tube and the inner tube are buried in the molten material in the vessel main body, accordingly, a fluid having its temperature elevated by absorption of the heat of the molten material rises along the inner circumferential surface of the outer tube, abuts against the lower surface of the lid and cooled by exchanging heat with the lid and forms a circulating flow. Since the heat in the molten material is continuously absorbed by the fluid, transferred to the lid and released from the lid to the atmospheric air, heat releasing efficiency can be improved compared with conventional cases. (N.H.)

ICRF heating on helical devices

International Nuclear Information System (INIS)

Rasmussen, D.A.; Lyon, J.F.; Hoffman, D.J.; Murakami, M.; England, A.C.; Wilgen, J.B.; Jaeger, E.F.; Wang, C.; Batchelor, D.B.

1995-01-01

Ion cyclotron range of frequency (ICRF) heating is currently in use on CHS and W7-AS and is a major element of the heating planned for steady state helical devices. In helical devices, the lack of a toroidal current eliminates both disruptions and the need for ICRF current drive, simplifying the design of antenna structures as compared to tokamak applications. However the survivability of plasma facing components and steady state cooling issues are directly applicable to tokamak devices. Results from LHD steady state experiments should be available on a time scale to strongly influence the next generation of steady state tokamak experiments. The helical plasma geometry provides challenges not faced with tokamak ICRF heating, including the potential for enhanced fast ion losses, impurity accumulation, limited access for antenna structures, and open magnetic field lines in the plasma edge. The present results and near term plans provide the basis for steady state ICRF heating of larger helical devices. An approach which includes direct electron, mode conversion, ion minority and ion Bernstein wave heating addresses these issues

ICRF heating on helical devices

International Nuclear Information System (INIS)

Rasmussen, D.A.; Lyon, J.F.; Hoffman, D.J.

1995-01-01

Ion cyclotron range of frequency (ICRF) heating is currently in use on CHS and W7AS and is a major element of the heating planned for steady state helical devices. In helical devices, the lack of a toroidal current eliminates both disruptions and the need for ICRF current drive, simplifying the design of antenna structures as compared to tokamak applications. However the survivability of plasma facing components and steady state cooling issues are directly applicable to tokamak devices. Results from LHD steady state experiments should be available on a time scale to strongly influence the next generation of steady state tokamak experiments. The helical plasma geometry provides challenges not faced with tokamak ICRF heating, including the potential for enhanced fast ion losses, impurity accumulation, limited access for antenna structures, and open magnetic field lines in the plasma edge. The present results and near term plans provide the basis for steady state ICRF heating of larger helical devices. An approach which includes direct electron, mode conversion, ion minority and ion Bernstein wave heating addresses these issues

Division of Development and Technology Plasma/Materials Interaction and High Heat Flux Materials and Components Task Groups: Report on the joint meeting, July 9, 1986

International Nuclear Information System (INIS)

Watson, R.D.

1986-09-01

This paper contains a collection of viewgraphs from a joint meeting of the Division of Development and Technology Plasma/Materials Interaction and High Heat Flux Materials and Components Task Groups. A list of contributing topics is: PPPL update, ATF update, Los Alamos RFP program update, status of DIII-D, PMI graphite studies at ORNL, PMI studies for low atomic number materials, high heat flux materials issues, high heat flux testing program, particle confinement in tokamaks, helium self pumping, self-regenerating coatings technical planning activity and international collaboration update

PCCS model development for SBWR using the CONTAIN code

International Nuclear Information System (INIS)

Tills, J.; Murata, K.K.; Washington, K.E.

1994-01-01

The General Electric Simplified Boiling Water Reactor (SBWR) employs a passive containment cooling system (PCCS) to maintain long-term containment gas pressure and temperature below design limits during accidents. This system consists of a steam supply line that connects the upper portion of the drywell with a vertical shell-and-tube single pass heat exchanger located in an open water pool outside of the containment safety envelope. The heat exchanger tube outlet is connected to a vent line that is submerged below the suppression pool surface but above the main suppression pool horizontal vents. Steam generated in the post-shutdown period flows into the heat exchanger tubes as the result of suction and/or a low pressure differential between the drywell and suppression chamber. Operation of the PCCS is complicated by the presence of noncondensables in the flow stream. Build-up of noncondensables in the exchanger and vent line for the periods when the vent is not cleared causes a reduction in the exchanger heat removal capacity. As flow to the exchanger is reduced due to the noncondensable gas build-up, the drywell pressure increases until the vent line is cleared and the noncondensables are purged into the suppression chamber, restoring the heat removal capability of the PCCS. This paper reports on progress made in modeling SBWR containment loads using the CONTAIN code. As a central part of this effort, a PCCS model development effort has recently been undertaken to implement an appropriate model in CONTAIN. The CONTAIN PCCS modeling approach is discussed and validated. A full SBWR containment input deck has also been developed for CONTAIN. The plant response to a postulated design basis accident (DBA) has been calculated with the CONTAIN PCCS model and plant deck, and the preliminary results are discussed

Heat transfer 1990. Proceedings of the ninth international heat transfer conference

International Nuclear Information System (INIS)

Hetsroni, G.

1990-01-01

This book contains the proceedings of the Ninth International Heat Transfer Conference. Included in Volume 3 are the following chapters: Refrigerant vapor condensation on a horizontal tube bundle. Local heat transfer in a reflux condensation inside a closed two-phase thermosyphon and surface temperature by means of a pulsed photothermal effects

Study of condensation heat transfer following a main steam line break inside containment

Energy Technology Data Exchange (ETDEWEB)

Cho, J.H.; Elia, F.A. Jr.; Lischer, D.J. [Stone & Webster Engineering Corporation, Boston, MA (United States)

1995-09-01

An alternative model for calculating condensation heat transfer following a main stream line break (MSLB) accident is proposed. The proposed model predictions and the current regulatory model predictions are compared to the results of the Carolinas Virginia Tube Reactor (CVTR) test. The very conservative results predicted by the current regulatory model result from: (1) low estimate of the condensation heat transfer coefficient by the Uchida correlation and (2) neglecting the convective contribution to the overall heat transfer. Neglecting the convection overestimates the mass of steam being condensed and does not permit the calculation of additional convective heat transfer resulting from superheated conditions. In this study, the Uchida correlation is used, but correction factors for the effects of convection an superheat are derived. The proposed model uses heat and mass transfer analogy methods to estimate to convective fraction of the total heat transfer and bases the steam removal rate on the condensation heat transfer portion only. The results predicted by the proposed model are shown to be conservative and more accurate than those predicted by the current regulatory model when compared with the results of the CVTR test. Results for typical pressurized water reactors indicate that the proposed model provides a basis for lowering the equipment qualification temperature envelope, particularly at later times following the accident.

Consumption of Dairy Yogurt Containing Lactobacillus paracasei ssp. paracasei, Bifidobacterium animalis ssp. lactis and Heat-Treated Lactobacillus plantarum Improves Immune Function Including Natural Killer Cell Activity

Science.gov (United States)

Lee, Ayoung; Lee, Young Ju; Yoo, Hye Jin; Kim, Minkyung; Chang, Yeeun; Lee, Dong Seog; Lee, Jong Ho

2017-01-01

The aim of this study was to investigate the impact of consuming dairy yogurt containing Lactobacillus paracasei ssp. paracasei (L. paracasei), Bifidobacterium animalis ssp. lactis (B. lactis) and heat-treated Lactobacillus plantarum (L. plantarum) on immune function. A randomized, open-label, placebo-controlled study was conducted on 200 nondiabetic subjects. Over a twelve-week period, the test group consumed dairy yogurt containing probiotics each day, whereas the placebo group consumed milk. Natural killer (NK) cell activity, interleukin (IL)-12 and immunoglobulin (Ig) G1 levels were significantly increased in the test group at twelve weeks compared to baseline. Additionally, the test group had significantly greater increases in serum NK cell activity and interferon (IFN)-γ and IgG1 than placebo group. Daily consumption of dairy yogurt containing L. paracasei, B. lactis and heat-treated L. plantarum could be an effective option to improve immune function by enhancing NK cell function and IFN-γ concentration (ClinicalTrials.gov: NCT03051425). PMID:28561762

Consumption of Dairy Yogurt Containing Lactobacillus paracasei ssp. paracasei, Bifidobacterium animalis ssp. lactis and Heat-Treated Lactobacillus plantarum Improves Immune Function Including Natural Killer Cell Activity

Directory of Open Access Journals (Sweden)

Ayoung Lee

2017-05-01

Full Text Available The aim of this study was to investigate the impact of consuming dairy yogurt containing Lactobacillus paracasei ssp. paracasei (L. paracasei, Bifidobacterium animalis ssp. lactis (B. lactis and heat-treated Lactobacillus plantarum (L. plantarum on immune function. A randomized, open-label, placebo-controlled study was conducted on 200 nondiabetic subjects. Over a twelve-week period, the test group consumed dairy yogurt containing probiotics each day, whereas the placebo group consumed milk. Natural killer (NK cell activity, interleukin (IL-12 and immunoglobulin (Ig G1 levels were significantly increased in the test group at twelve weeks compared to baseline. Additionally, the test group had significantly greater increases in serum NK cell activity and interferon (IFN-γ and IgG1 than placebo group. Daily consumption of dairy yogurt containing L. paracasei, B. lactis and heat-treated L. plantarum could be an effective option to improve immune function by enhancing NK cell function and IFN-γ concentration (ClinicalTrials.gov: NCT03051425.

Consumption of Dairy Yogurt Containing Lactobacillus paracasei ssp. paracasei, Bifidobacterium animalis ssp. lactis and Heat-Treated Lactobacillus plantarum Improves Immune Function Including Natural Killer Cell Activity.

Science.gov (United States)

Lee, Ayoung; Lee, Young Ju; Yoo, Hye Jin; Kim, Minkyung; Chang, Yeeun; Lee, Dong Seog; Lee, Jong Ho

2017-05-31

The aim of this study was to investigate the impact of consuming dairy yogurt containing Lactobacillus paracasei ssp. paracasei ( L. paracasei ), Bifidobacterium animalis ssp. lactis ( B. lactis ) and heat-treated Lactobacillus plantarum ( L. plantarum ) on immune function. A randomized, open-label, placebo-controlled study was conducted on 200 nondiabetic subjects. Over a twelve-week period, the test group consumed dairy yogurt containing probiotics each day, whereas the placebo group consumed milk. Natural killer (NK) cell activity, interleukin (IL)-12 and immunoglobulin (Ig) G1 levels were significantly increased in the test group at twelve weeks compared to baseline. Additionally, the test group had significantly greater increases in serum NK cell activity and interferon (IFN)-γ and IgG1 than placebo group. Daily consumption of dairy yogurt containing L. paracasei , B. lactis and heat-treated L. plantarum could be an effective option to improve immune function by enhancing NK cell function and IFN-γ concentration (ClinicalTrials.gov: NCT03051425).

The effect of external boundary conditions on condensation heat transfer in rotating heat pipes

Science.gov (United States)

Daniels, T. C.; Williams, R. J.

1979-01-01

Experimental evidence shows the importance of external boundary conditions on the overall performance of a rotating heat pipe condenser. Data are presented for the boundary conditions of constant heat flux and constant wall temperature for rotating heat pipes containing either pure vapor or a mixture of vapor and noncondensable gas as working fluid.

Heat transfer studies in pool fire environment

International Nuclear Information System (INIS)

Nitsche, F.

1993-01-01

A Type B package has to withstand severe thermal accident conditions. To calculate the temperature behaviour of such a package in a real fire environment, heat transfer parameters simulating the effect of the fire are needed. For studying such heat transfer parameters, a systematic programme of experimental and theoretical investigations was performed which was part of the IAEA Coordinated Research Programme (Nitsche and Weib 1990). The studies were done by means of small, unfinned and finned, steel model containers of simplified design in hydrocarbon fuel open fire tests. By using various methods, flame and container temperatures were measured and also container surface absorptivity before and after the test to study the effect of sooting and surface painting on heat transfer. Based on all these experimental data and comparative calculations, simplified, effective heat transfer parameters could be derived, simulating the effect of the real fire on the model containers. (J.P.N.)

Reactor waste heat utilization and district heating reactors. Nuclear district heating in Sweden - Regional reject heat utilization schemes and small heat-only reactors

International Nuclear Information System (INIS)

Hannerz, K.; Larsson, Y.; Margen, P.

1977-01-01

A brief review is given of the current status of district heating in Sweden. In future, district heating schemes will become increasingly interesting as a means of utilizing heat from nuclear reactors. Present recommendations in Sweden are that large reactors should not be located closer than about 20 km from large population centres. Reject heat from such reactors is cheap at source. To minimize the cost of long distance hot water transmission large heat rates must be transmitted. Only areas with large populations can meet this requirement. The three areas of main interest are Malmoe/Lund/Helsingborg housing close to 0.5 million; Greater Stockholm housing 1 to 1.5 million and Greater Gothenburg housing about 0.5 million people. There is an active proposal that the Malmoe/Lund/Helsingborg region would be served by a third nuclear unit at Barsebaeck, located about 20 km from Malmoe/Lund and supplying 950 MW of base load heat. Preliminary proposals for Stockholm involve a 2000 MW heat supply; proposals for Gothenburg are more tentative. The paper describes progress on these proposals and their technology. It also outlines technology under development to increase the economic range of large scale heat transport and to make distribution economic even for low heat-density family housing estates. Regions apart from the few major urban areas mentioned above require the adoption of a different approach. To this end the development of a small, simple low-temperature reactor for heat-only production suitable for urban location has been started in Sweden in close contact with Finland. Some results of the work in progress are presented, with emphasis on the safety requirements. An outline is given in the paper as to how problems of regional heat planning and institutional and legislative issues are being approached

Considerations in modelling the melting of fuel containing fission products and solute oxides

International Nuclear Information System (INIS)

Akbari, F.; Welland, M.J.; Lewis, B.J.; Thompson, W.T.

2005-01-01

It is well known that the oxidation of a defected fuel element by steam gives rise to an increase in O/U ratio with a consequent lowering of the incipient melting temperature. Concurrently, the hyperstoichiometry reduces the thermal conductivity thereby raising the centerline fuel pellet temperature for a fixed linear power. The development of fission products soluble in the UO 2 phase or, more important, the deliberate introduction of additive oxides in advanced CANDU fuel bundle designs further affects and generally lowers the incipient melting temperature. For these reasons, the modeling of the molten (hyperstoichiometric) UO 2 phase containing several solute oxides (ZrO 2 , Ln 2 O 3 and AnO 2 ) is advancing in the expectation of developing a moving boundary heat and mass transfer model aimed at better defining the limits of safe operating practice as burnup advances. The paper describes how the molten phase stability model is constructed. The redistribution of components across the solid-liquid interface that attends the onset of melting of a non-stoichiometric UO 2 containing several solutes will be discussed. The issues of how to introduce boundary conditions into heat transfer calculations consistent with the requirements of the Phase Rule will be addressed. The Stefan problem of a moving boundary associated with the solid/liquid interface sets this treatment apart from conventional heat and mass transfer problems. (author)

Convection heat transfer

CERN Document Server

Bejan, Adrian

2013-01-01

Written by an internationally recognized authority on heat transfer and thermodynamics, this second edition of Convection Heat Transfer contains new and updated problems and examples reflecting real-world research and applications, including heat exchanger design. Teaching not only structure but also technique, the book begins with the simplest problem solving method (scale analysis), and moves on to progressively more advanced and exact methods (integral method, self similarity, asymptotic behavior). A solutions manual is available for all problems and exercises.

Investigations on passive containment cooling; Untersuchungen zur passiven Containmentkuehlung

Energy Technology Data Exchange (ETDEWEB)

Knebel, J.U.; Cheng, X.; Neitzel, H.J.; Erbacher, F.J. [Forschungszentrum Karlsruhe GmbH Technik und Umwelt (Germany). Inst. fuer Angewandte Thermo- und Fluiddynamik; Hofmann, F. [Forschungszentrum Karlsruhe GmbH Technik und Umwelt (Germany). Projekt Nukleare Sicherheitsforschung

1997-12-31

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) [Deutsch] Das im PASCO Programm untersuchte Verbundcontainment fuer zukuenftige Leichtwasserreaktoren ist ein erfolgversprechendes rein passives System zur Nachwaermeabfuhr nach einem schweren Stoerfall. Die passiven Mechanismen der Waermeuebertragung sind Naturkonvektion und Waermestrahlung. Die durch Waermestrahlung uebertragene Waerme betraegt fuer einen Emissionskoeffizienten von 0.9 etwa 50% der insgesamt ueber das Stahl-Containment abgefuehrten Waerme und ist somit von entscheidender Bedeutung fuer die Nachwaermeabfuhr. Fuer die waermeuebertragenden Oberflaechen ist ein hoher Emissionkoeffizient erforderlich. Das im SUCO Programm untersuchte Sumpfkuehlkonzept realisiert eine stabile Naturkonvektionsstroemung zwischen Waermequelle und Waermesenke. (orig./DG)

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.)

GOTHIC Simulation of Passive Containment Cooling System

International Nuclear Information System (INIS)

Ha, Huiun; Kim, Hangon

2013-01-01

The performance of this system depends on the condensation of steam moving downward inside externally cooled vertical tubes. AES-2006: During a DBA, heat is removed by internally cooled vertical tubes, which are located in containment. We are currently developing the conceptual design of Innovative PWR, which is will be equipped with various passive safety features, including PCCS. We have plan to use internal heat exchanger (HX) type PCCS with concrete containment. In this case, the elevation of HXs is important to ensure the heat removal during accidents. In general, steam is lighter than air mixture in containment. So, steam may be collected at the upper side of containment. It means that higher elevation of HXs, larger heat removal efficiency of those. So, the aim of the present paper is to give preliminary study on variation of heat removal performance according to elevation of HXs. With reference to the design specification of the current reactors including APR+, we had determined conceptual design of PCCS. Using it, we developed a GOTHIC model of the APR1400 containment was adopted PCCS. This calculation model is described herein and representative results of calculation are presented. APR 1400 GOTHIC model was developed for PCCS performance calculation and sensitivity test according to installation elevation of PCCXs. Calculation results confirm that PCCS is working properly. It is found that the difference due to the installation elevation of PCCXs is insignificant at this preliminary analysis, however, further studies should be performed to confirm final performance of PCCS according to the installation elevation. These insights are important for developing the PCCS of Innovative PWR

GOTHIC Simulation of Passive Containment Cooling System

Energy Technology Data Exchange (ETDEWEB)

Ha, Huiun; Kim, Hangon [Korea Hydro and Nuclear Power Co. Ltd., Daejeon (Korea, Republic of)

2013-05-15

The performance of this system depends on the condensation of steam moving downward inside externally cooled vertical tubes. AES-2006: During a DBA, heat is removed by internally cooled vertical tubes, which are located in containment. We are currently developing the conceptual design of Innovative PWR, which is will be equipped with various passive safety features, including PCCS. We have plan to use internal heat exchanger (HX) type PCCS with concrete containment. In this case, the elevation of HXs is important to ensure the heat removal during accidents. In general, steam is lighter than air mixture in containment. So, steam may be collected at the upper side of containment. It means that higher elevation of HXs, larger heat removal efficiency of those. So, the aim of the present paper is to give preliminary study on variation of heat removal performance according to elevation of HXs. With reference to the design specification of the current reactors including APR+, we had determined conceptual design of PCCS. Using it, we developed a GOTHIC model of the APR1400 containment was adopted PCCS. This calculation model is described herein and representative results of calculation are presented. APR 1400 GOTHIC model was developed for PCCS performance calculation and sensitivity test according to installation elevation of PCCXs. Calculation results confirm that PCCS is working properly. It is found that the difference due to the installation elevation of PCCXs is insignificant at this preliminary analysis, however, further studies should be performed to confirm final performance of PCCS according to the installation elevation. These insights are important for developing the PCCS of Innovative PWR.

FFTF primary heat transport system heating, ventilating and air conditioning system experience

International Nuclear Information System (INIS)

Umek, A.M.; Hicks, D.F.; Schweiger, D.L.

1981-01-01

FFTF cools its primary/in-containment sodium equipment cells by means of a forced nitrogen cooling system which exchanges heat with a water-glycol system. The nitrogen cooling system is also used to maintain an inert gas atmosphere in the cells containing sodium equipment. Sodium Piping and Components have installed electrical resistance heaters to maintain a minimum sodium temperature and stainless steel jacketed mineral insulation to reduce heat loss. Design features and test results of a comprehensive redesign of the HVAC and insulation system required to support long-term nuclear operations are discussed

Device for separating, purifying and recovering nuclear fuel material, impurities and materials from impurity-containing nuclear fuel materials or nuclear fuel containing material

International Nuclear Information System (INIS)

Sato, Ryuichi; Kamei, Yoshinobu; Watanabe, Tsuneo; Tanaka, Shigeru.

1988-01-01

Purpose: To separate, purify and recover nuclear fuel materials, impurities and materials with no formation of liquid wastes. Constitution: Oxidizing atmosphere gases are introduced from both ends of a heating furnace. Vessels containing impurity-containing nuclear fuel substances or nuclear fuel substance-containing material are continuously disposed movably from one end to the other of the heating furnace. Then, impurity oxides or material oxides selectively evaporated from the impurity-containing nuclear fuel substances or nuclear fuel substance-containing materials are entrained in the oxidizing atmosphere gas and the gases are led out externally from a discharge port opened at the intermediate portion of the heating furnace, filters are disposed to the exit to solidify and capture the nuclear fuel substances and traps are disposed behind the filters to solidify and capture the oxides by spontaneous air cooling or water cooling. (Sekiya, K.)

A non-equilibrium model for soil heating and moisture transport during extreme surface heating: The soil (heat-moisture-vapor) HMV-Model Version

Science.gov (United States)

William Massman

2015-01-01

Increased use of prescribed fire by land managers and the increasing likelihood of wildfires due to climate change require an improved modeling capability of extreme heating of soils during fires. This issue is addressed here by developing and testing the soil (heat-moisture-vapor) HMVmodel, a 1-D (one-dimensional) non-equilibrium (liquid- vapor phase change)...

Heat transfer--Orlando (Symposium), 1980

International Nuclear Information System (INIS)

Stein, R.P.

1980-01-01

This conference proceedings contains 36 papers of which 3 appear as abstracts. 23 papers are indexed separately. Topics covered include: thermodynamics of PWR and LMFBR Steam Generators; two-phase flow in parallel channels; geothermal heat transfer; natural circulation in complex geometries; heat transfer in non-Newtonian systems; and process heat transfer

Utilization of ruthenium volatilization at heating of residue containing phosphates and nitrates for ruthenium separation and for its qualitative proof

International Nuclear Information System (INIS)

Holgye, Z.

1979-01-01

The volatility of ruthenium during the heating of a residue after evaporation of a solution containing ruthenium, phosphates and nitrates may be utilized for the separation of ruthenium from various substances. Sup(103,106) Ru may be rapidly, selectively, and quantitatively separated from fission products mixture. Ruthenium may be also separated in this way from various inorganic salts or from biological material. The volatility of ruthenium may be used also for its qualitative proof. (author)

Double containment shell for nuclear power plants

International Nuclear Information System (INIS)

Sykora, D.

1977-01-01

A double containment shell is proposed for nuclear power plants, especially those equipped with pressurized water reactors. The shell offers increased environmental protection from primary circuit accidents. The inner shell is built of steel or concrete while the outer shell is always built of concrete. The space between the two shells is filled with water and is provided with several manholes and with stiffeners designed for compensation for load due to the water hydrostatic pressure. Water serves the airtight separation of the containment shell inside from the environment and the absorption of heat released in a primary circuit accident. In case the inner shell is made of concrete, it is provided with heat-removal tubes in-built in its walls ensuring rapid heat transfer from the inside of the containment to the water in the interwall space. (Z.M.)

Solar heating system

Science.gov (United States)

Schreyer, James M.; Dorsey, George F.

1982-01-01

An improved solar heating system in which the incident radiation of the sun is absorbed on collector panels, transferred to a storage unit and then distributed as heat for a building and the like. The improvement is obtained by utilizing a storage unit comprising separate compartments containing an array of materials having different melting points ranging from 75.degree. to 180.degree. F. The materials in the storage system are melted in accordance with the amount of heat absorbed from the sun and then transferred to the storage system. An efficient low volume storage system is provided by utilizing the latent heat of fusion of the materials as they change states in storing and releasing heat for distribution.

[Depyrogenation test regarding inox and glass containers in the preparation of parenteral nutrition mixtures].

Science.gov (United States)

Lajoinie, A; Vasselon, P; Tall, M-L; Salmon, D; Bréant, V; Diouf, E; Pivot, C; Pirot, F

2012-09-01

The preparation of parenteral nutrition mixture (PNM) in an open chamber requires the use of intermediate containers sterile and non-pyrogenic. A sterilization of containers by moist heat in large autoclaves is the suitable method. However, sterilization by moist heat is not a depyrogenation method. In our study, we report the validation of a sterilization and depyrogenation method for containers by dry heat using a convection oven. Sterilization and depyrogenation of material by dry heat have been audited by the reduction of at least three logarithms of original endotoxin rate. The containers were initially artificially contaminated with a suspension of endotoxin for 16 hours. Contaminated containers were placed in an oven with revolving heat at 250 °C for 1 hour. After treatment with dry heat, the residual endotoxin levels in the containers were determined by a kinetic chromogenic method. After treatment with dry heat, the average log reductions of endotoxin levels were respectively, for glass and steel containers, 4.78 ± 0.07 and 4.87 ± 0.03. The present validation study confirms the effectiveness of treatment with dry heat for sterilization and depyrogenation of glass and steel containers. This method of sterilization and depyrogenation meets the microbiological quality requirements for the preparation of MNP. Copyright © 2012 Elsevier Masson SAS. All rights reserved.

Investigation on chemical heat pump using calcium-chloride; Enka calcium no suiwa dassui hanno wo mochiita solar chemical heat pump ni kansuru kenkyu

Energy Technology Data Exchange (ETDEWEB)

Fujii, I; Arai, T; Saito, Y [Meiji University, Tokyo (Japan)

1997-11-25

With an objective of developing a room heating system utilizing a solar chemical heat pump, an experimental system was fabricated to evaluate its performance. Steam was employed as a working gas, and for a reaction material, calcium-chloride was used, which has a reaction temperature zone permitting safe use and fitting the purpose among other hydrate systems and has high standard enthalpy in hydration. Water was used as a solar heat transferring medium. The system operates under the following principle: a container I is filled with hydrated salt and a container II with water, the two containers being linked with a pipe interposed with a valve; heat is inputted and outputted by performing charging and discharging alternately; and the role of a heat pump is played by deriving from environment the heat of water evaporation in the container II during discharging. The COP must take into account the electric power consumption of the water circulation pump to transfer solar heat. A COP of 0.256 was derived as a result of the experiment. 3 refs., 5 figs.

Heat and water transfer in a rotating drum containing solid substrate particles

NARCIS (Netherlands)

Schutyser, M.A.I.; Weber, F.J.; Briels, W.J.; Rinzema, A.; Boom, R.M.

2003-01-01

In previous work we reported on the simulation of mixing behavior of a slowly rotating drum for solid-state fermentation (SSF) using a discrete particle model. In this investigation the discrete particle model is extended with heat and moisture transfer. Heat transfer is implemented in the model via

Review of EU-APR Design for Selected Safety Issues of WERNA RHWG 2013

Energy Technology Data Exchange (ETDEWEB)

Kim, Yong Soo; Kim, Ji Hwan [KHNP CRI, Daejeon (Korea, Republic of)

2016-10-15

Western European Nuclear Regulators' Association (WENRA) was established in 1999 to develop a harmonized approach to nuclear safety and radiation protection and their regulation. In 2013, the Reactor Harmonization Working Group (RHWG) of WENRA sets out the common positions on the seven selected key safety issues. This paper is to introduce the regulatory positions of WENRA RHWG 2013 and to review the compliance of the EU-APR with them. In this paper, we reviewed the compliance of the EUAPR regarding seven safety issues for new NPPs presented by WERNA RHWG in 2013. The EU-APR design fully complies with all WERNA RHWG safety issues since the following measures have been incorporated in it: - Successive five levels of DiD maintaining independence between different levels of DiD - Diverse design against multiple failure events such as ATWS, SBO, Loss of Ultimate Heat Sink, and Loss of Spent Fuel Pool Cooling - SAs dedicated mitigation systems to ensure the containment integrity during the SAs. - Practically eliminates accident sequences with a large or early release of radiological materials by diverse designs for multiple failure events, SAs dedicated mitigation system, and double containment design - Standard site parameters not lead to core melt accidents due to natural or man-made external hazards.

Regulatory analysis for the resolution of Unresolved Safety Issue A-44, Station Blackout. Draft report

International Nuclear Information System (INIS)

Rubin, A.M.

1986-01-01

''Station Blackout'' is the complete loss of alternating current (ac) electric power to the essential and nonessential buses in a nuclear power plant; it results when both offsite power and the onsite emergency ac power systems are unavailable. Because many safety systems required for reactor core decay heat removal and containment heat removal depend on ac power, the consequences of a station blackout could be severe. Because of the concern about the frequency of loss of offsite power, the number of failures of emergency diesel generators, and the potentially severe consequences of a loss of all ac power, ''Station Blackout'' was designated as Unresolved Safety Issue (USI) A-44. This report presents the regulatory analysis for USI A-44. It includes: (1) a summary of the issue, (2) the proposed technical resolution, (3) alternative resolutions considered by the Nuclear Regulatory Commission (NRC) staff, (4) an assessment of the benefits and costs of the recommended resolution, (5) the decision rationale, and (6) the relationship between USI A-44 and other NRC programs and requirements

Heat and thermodynamics

CERN Document Server

Saxena, A K

2014-01-01

Heat and thermodynamics aims to serve as a textbook for Physics, Chemistry and Engineering students. The book covers basic ideas of Heat and Thermodynamics, Kinetic Theory and Transport Phenomena, Real Gases, Liquafaction and Production and Measurement of very Low Temperatures, The First Law of Thermodynamics, The Second and Third Laws of Thermodynamics and Heat Engines and Black Body Radiation. KEY FEATURES Emphasis on concepts Contains 145 illustrations (drawings), 9 Tables and 48 solved examples At the end of chapter exercises and objective questions

Active heat exchange system development for latent heat thermal energy storage

Science.gov (United States)

Alario, J.; Kosson, R.; Haslett, R.

1980-01-01

Various active heat exchange concepts were identified from among three generic categories: scrapers, agitators/vibrators and slurries. The more practical ones were given a more detailed technical evaluation and an economic comparison with a passive tube-shell design for a reference application (300 MW sub t storage for 6 hours). Two concepts were selected for hardware development: (1) a direct contact heat exchanger in which molten salt droplets are injected into a cooler counterflowing stream of liquid metal carrier fluid, and (2) a rotating drum scraper in which molten salt is sprayed onto the circumference of a rotating drum, which contains the fluid salt is sprayed onto the circumference of a rotating drum, which contains the fluid heat sink in an internal annulus near the surface. A fixed scraper blade removes the solidified salt from the surface which was nickel plated to decrease adhesion forces. In addition to improving performance by providing a nearly constant transfer rate during discharge, these active heat exchanger concepts were estimated to cost at least 25% less than the passive tube-shell design.

Liquid Salt Heat Exchanger Technology for VHTR Based Applications

Energy Technology Data Exchange (ETDEWEB)

Anderson, Mark; Sridhara, Kumar; Allen, Todd; Peterson, Per

2012-10-11

The objective of this research is to evaluate performance of liquid salt fluids for use as a heat carrier for transferring high-temperature process heat from the very high-temperature reactor (VHTR) to chemical process plants. Currently, helium is being considered as the heat transfer fluid; however, the tube size requirements and the power associated with pumping helium may not be economical. Recent work on liquid salts has shown tremendous potential to transport high-temperature heat efficiently at low pressures over long distances. This project has two broad objectives: To investigate the compatibility of Incoloy 617 and coated and uncoated SiC ceramic composite with MgCl2-KCl molten salt to determine component lifetimes and aid in the design of heat exchangers and piping; and, To conduct the necessary research on the development of metallic and ceramic heat exchangers, which are needed for both the helium-to-salt side and salt-to-process side, with the goal of making these heat exchangers technologically viable. The research will consist of three separate tasks. The first task deals with material compatibility issues with liquid salt and the development of techniques for on-line measurement of corrosion products, which can be used to measure material loss in heat exchangers. Researchers will examine static corrosion of candidate materials in specific high-temperature heat transfer salt systems and develop an in situ electrochemical probe to measure metallic species concentrations dissolved in the liquid salt. The second task deals with the design of both the intermediate and process side heat exchanger systems. Researchers will optimize heat exchanger design and study issues related to corrosion, fabrication, and thermal stresses using commercial and in-house codes. The third task focuses integral testing of flowing liquid salts in a heat transfer/materials loop to determine potential issues of using the salts and to capture realistic behavior of the salts in a

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.)

Heat transfer and mechanical interactions in fusion nuclear systems

International Nuclear Information System (INIS)

Nygren, R.E.

1984-01-01

This general review of design issues in heat transfer and mechanical interactions of the first wall, blanket and shield systems of tokamak and mirror fusion reactors begins with a brief introduction to fusion nuclear systems. The design issues are summarized in tables and the following examples are described to illustrate these concerns: the surface heating of limiters, heat transfer from solid breeders, MHD effects in liquid metal blankets, mechanical loads from electromagnetic transients and remote maintenance

CASTOR registered HAW28M - a high heat load cask for transport and storage of vitrified high level waste containers

International Nuclear Information System (INIS)

Vossnacke, A.; Klein, K.; Kuehne, B.

2004-01-01

Within the German return programme for vitrified high level waste (HLW) from reprocessing at COGEMA and BNFL up to now 39 casks loaded with 28 containers each were transported back to Germany and are stored in the Interim Storage Facility Gorleben (TBL-G) for up to 40 years. For transport and storage in all but one case the GNB casks CASTOR registered HAW 20/28 CG have been used. This cask type is designed to accommodate 20 or 28 HLW containers with a total thermal power of 45 kW maximum. In the near future, among the high level waste, which has to be returned to Germany, there will be an increasing number of containers of which the heat capacity and radioactive inventory will exceed the technical limits of the CASTOR registered HAW 20/28 CG. Therefore GNB has started the development of a new cask generation, named CASTOR registered HAW28M, meeting these future requirements. The CASTOR registered HAW28M is especially developed for the transport of vitrified residues from France and Great Britain to Germany. It complies with the international regulations for type B packages according to IAEA (International Atomic Energy Agency). It is thus guaranteed that even in case of any accident the cask body and the lid system remain functional and the safe confinement of the radioactive contents remains intact during transport. The CASTOR registered HAW28M fulfills not only the requirements for transport but also the acceptance criteria of interim storage: radiation shielding, heat dissipation, safe confinement under both normal and hypothetical accident conditions. Storage buildings such as the TBL-G simply support the safety functions of the cask. The challenge for the development results from higher requirements of the technical specification, particularly related to fuel which is reprocessed. As a consequence of the reprocessing of fuel with increased enrichment and burn up, higher heat capacity and sophisticated shielding measures have to be considered. For the CASTOR

VII international district heating conference

International Nuclear Information System (INIS)

1988-01-01

The proceedings of the 7th International District Heating Conference contain the full texts of the 89 presented papers of which 11 fall under the INIS Subject Scope. The conference met in seven sessions and dealt with the following problem areas: design and optimization of systems of district heating, integration of the power system and the district heating systems, cooperation of nuclear and fossil burning sources in district heating systems, the use of specific nuclear power plants for heating purposes, questions of the control of systems of district heating, the development of components of heating networks, the reliability and design of heat supply pipes. (Z.M.)

Nuclear reactor container

International Nuclear Information System (INIS)

Fujimoto, Kiyoshi; Kataoka, Yoshiyuki; Murase, Michio; Fujii, Tadashi; Susuki, Akira.

1994-01-01

A wet well space above a pressure suppression pool is divided into a first wet well on the side in contact with the pressure suppression pool and a second wet well on the side not in contact with the pool. Cooling water is contained in the second wet well and it is in communication with the first wet well by pipelines. Since steams flown into the second well are condensed in the cooling water, they continuously transfer from the first wet well to the second wet well, thereby capable of eliminating the effects of incondensible gases in the first wet well. With such procedures, the effect of the incondensible gases can be eliminated even without cooling from the outside of the reactor. Heat accumulation can be increased in a container of any material, so that thermal load on cooling circuits for removing after-heat can be mitigated. (T.M.)

Feasibility Study to Reduce Thermal Resistance of Finned Containment Wall in Simplified OPR1000

Energy Technology Data Exchange (ETDEWEB)

Lee, Jong Hwi; Kang, Hie Chan [Kunsan National University, Gunsan (Korea, Republic of); Noh, Hyung Gyun [Pohang University, Pohang (Korea, Republic of)

2016-05-15

This concept is securing of cooling capability by using finned containment itself, it could be another alternative for achieving decompression of containment as heat sink. The objective of this study is a feasibility test to estimate the heat transfer performance from the finned containment wall in case of OPR1000. The commercial code, ANSYS CFX 16 was used in this work. The number of grids is about 1.8 million. Therefore, 250mm rebar affects more considerable than that of 50mm to the temperature distribution. For this reasons, temperature distribution of z-axis direction was showed significant changes in (c). The heat transfer in three types of containment was 267.6W, 265.2W and 307.8W, respectively. The Type B case increased up to 15% of heat transfer than the baseline containment building. Three different types of containment wall were tested by numerical simulation to understand the cooling performance of finned containment wall. We can conclude as follows: For the finned containment wall type A that fins are installed inside and outside with the same rebar configuration of conventional containment building, the heat transfer is almost the same as conventional containment wall. The finned containment wall type B that volume fraction of rebar is increased transfer the heat 15% more compared with conventional one. The cross-sectional area or volume fraction of the rebar to attach fin is important to enhance the heat transfer. The fin efficiency of the fin is very low as 3.1% in the present cases.

Global Issues

Energy Technology Data Exchange (ETDEWEB)

Seitz, J.L.

2001-10-15

Global Issues is an introduction to the nature and background of some of the central issues - economic, social, political, environmental - of modern times. This new edition of this text has been fully updated throughout and features expanded sections on issues such as global warming, biotechnology, and energy. Fully updated throughout and features expanded sections on issues such as global warming, biotechnology, and energy. An introduction to the nature and background of some of the central issues - economic, social, political, environmental - of modern times. Covers a range of perspectives on a variety of societies, developed and developing. Extensively illustrated with diagrams and photographs, contains guides to further reading, media, and internet resources, and includes suggestions for discussion and studying the material. (author)

Brown pigment formation in heated sugar-protein mixed suspensions containing unmodified and peptically modified whey protein concentrates.

Science.gov (United States)

Rongsirikul, Narumol; Hongsprabhas, Parichat

2016-01-01

Commercial whey protein concentrate (WPC) was modified by heating the acidified protein suspensions (pH 2.0) at 80 °C for 30 min and treating with pepsin at 37 °C for 60 min. Prior to spray-drying, such modification did not change the molecular weights (MWs) of whey proteins determined by sodium dodecyl sulfate-polyacrylamide gel electrophoresis (SDS-PAGE). After spray-drying the modified whey protein concentrate with trehalose excipient (MWPC-TH), it was found that the α-lactalbumin (α-La) was the major protein that was further hydrolyzed the most. The reconstituted MWPC-TH contained β-lactoglobulin (β-Lg) as the major protein and small molecular weight (MW) peptides of less than 6.5 kDa. The reconstituted MWPC-TH had higher NH2 group, Trolox equivalent antioxidant capacity (TEAC), lower exposed aromatic ring and thiol (SH) contents than did the commercial WPC. Kinetic studies revealed that the addition of MWPC-TH in fructose-glycine solution was able to reduce brown pigment formation in the mixtures heated at 80 to 95 °C by increasing the activation energy (Ea) of brown pigment formation due to the retardation of fluoresced advanced glycation end product (AGEs) formation. The addition of MWPC to reducing sugar-glycine/commercial WPC was also able to lower brown pigment formation in the sterilized (121 °C, 15 min) mixed suspensions containing 0.1 M reducing sugar and 0.5-1.0 % glycine and/or commercial (P < 0.05). It was demonstrated that the modification investigated in this study selectively hydrolyzed α-La and retained β-Lg for the production of antibrowning whey protein concentrate.

Container corporation: The financial, management, and operating implications of a universal beverage container recovery system: Executive summary

Energy Technology Data Exchange (ETDEWEB)

1991-01-01

This report contains an overview of the system. It discusses containers and container materials, elements of the system, and the container recovery cost structure. It includes a financial evaluation and addresses issues of implementation.

Experimental results of direct containment heating by high-pressure melt ejection into the Surtsey vessel: The DCH-3 and DCH-4 tests

International Nuclear Information System (INIS)

Allen, M.D.; Pilch, M.; Brockmann, J.E.; Tarbell, W.W.; Nichols, R.T.; Sweet, D.W.

1991-08-01

Two experiments, DCH-3 and DCH-4, were performed at the Surtsey test facility to investigate phenomena associated with a high-pressure melt ejection (HPME) reactor accident sequence resulting in direct containment heating (DCH). These experiments were performed using the same experimental apparatus with identical initial conditions, except that the Surtsey test vessel contained air in DCH-3 and argon in DCH-4. Inerting the vessel with argon eliminated chemical reactions between metallic debris and oxygen. Thus, a comparison of the pressure response in DCH-3 and DCH-4 gave an indication of the DCH contribution due to metal/oxygen reactions. 44 refs., 110 figs., 43 tabs

Experimental results of direct containment heating by high-pressure melt ejection into the Surtsey vessel: The DCH-3 and DCH-4 tests

Energy Technology Data Exchange (ETDEWEB)

Allen, M.D.; Pilch, M.; Brockmann, J.E.; Tarbell, W.W. (Sandia National Labs., Albuquerque, NM (United States)); Nichols, R.T. (Ktech Corp., Albuquerque, NM (United States)); Sweet, D.W. (AEA Technology, Winfrith (United Kingdom))

1991-08-01

Two experiments, DCH-3 and DCH-4, were performed at the Surtsey test facility to investigate phenomena associated with a high-pressure melt ejection (HPME) reactor accident sequence resulting in direct containment heating (DCH). These experiments were performed using the same experimental apparatus with identical initial conditions, except that the Surtsey test vessel contained air in DCH-3 and argon in DCH-4. Inerting the vessel with argon eliminated chemical reactions between metallic debris and oxygen. Thus, a comparison of the pressure response in DCH-3 and DCH-4 gave an indication of the DCH contribution due to metal/oxygen reactions. 44 refs., 110 figs., 43 tabs.

High pressure melt ejection (HPME) and direct containment heating (DCH): state-of-the-art report

International Nuclear Information System (INIS)

1996-12-01

This report first address the accident considerations leading to conditions with the reactor pressure vessel at a significant pressure. It also address those accident management actions that could prevent such a pressurized state and the effectiveness of operator actions since this is a principal focus of how a HPME could be prevented. Furthermore, it also investigates those situations, while very unlikely, in which the RCS could be at a significant pressure and possibly experience RPV failure. This represents a significant set of experimental information that, coupled with the integral effects models, provides the necessary insights for issue resolution for a number of containment types. Lastly, conclusions and recommendations are developed to be presented to the CSNI

Dynamic Complexity Study of Nuclear Reactor and Process Heat Application Integration

International Nuclear Information System (INIS)

Taylor, J'Tia Patrice; Shropshire, David E.

2009-01-01

This paper describes the key obstacles and challenges facing the integration of nuclear reactors with process heat applications as they relate to dynamic issues. The paper also presents capabilities of current modeling and analysis tools available to investigate these issues. A pragmatic approach to an analysis is developed with the ultimate objective of improving the viability of nuclear energy as a heat source for process industries. The extension of nuclear energy to process heat industries would improve energy security and aid in reduction of carbon emissions by reducing demands for foreign derived fossil fuels. The paper begins with an overview of nuclear reactors and process application for potential use in an integrated system. Reactors are evaluated against specific characteristics that determine their compatibility with process applications such as heat outlet temperature. The reactor system categories include light water, heavy water, small to medium, near term high-temperature, and far term high temperature reactors. Low temperature process systems include desalination, district heating, and tar sands and shale oil recovery. High temperature processes that support hydrogen production include steam reforming, steam cracking, hydrogen production by electrolysis, and far-term applications such as the sulfur iodine chemical process and high-temperature electrolysis. A simple static matching between complementary systems is performed; however, to gain a true appreciation for system integration complexity, time dependent dynamic analysis is required. The paper identifies critical issues arising from dynamic complexity associated with integration of systems. Operational issues include scheduling conflicts and resource allocation for heat and electricity. Additionally, economic and safety considerations that could impact the successful integration of these systems are considered. Economic issues include the cost differential arising due to an integrated system

MDEP Common Position CP-EPRWG-04. Common position on EPR containment heat removal system in accident conditions

International Nuclear Information System (INIS)

2015-01-01

the active parts; - containment heat removal, including corium cooling, during core melt accidents shall be provided; - it shall be possible to reduce containment pressure in a controlled manner in the long term taking into account the impact of non-condensable gases; - there shall be provisions to reduce the amount of fission products in the containment atmosphere in case of the core melt accident

Method for pre-heating lmfbr type reactors

International Nuclear Information System (INIS)

Yokozawa, Atsushi; Kataoka, Hajime.

1978-01-01

Purpose: To enable pre-heating for the inside of the reactor container and the inside of the coolant recycling system with no additional facilities. Method: The coolant recycling system is composed of a heat exchanger, a mechanical pump, a check valve, a flow meter or the like and it is connected in series by way of a pipe line to a reactor container. The mechanical pump is used as a gas recycling device upon pre-heating and it is designed so that a blower such as a fan can be replaced for the impeller of the pump. The inside of the reactor container and the inside of the coolant recycling system is at first filled with an inert gas such as for use with cover gas. Then, nuclear fuels are loaded to attain criticality. Simultaneously, the blower is started and the control rods are operated while cooling the nuclear fuel with the inert gas thus to obtain heat required for pre-heating the pipe line or the like from the nuclear fuels. After the completion of the pre-heating, the liquid metal is charged. (Ikeda, J.)

Loop heat pipes - highly efficient heat-transfer devices for systems of sun heat supply

Energy Technology Data Exchange (ETDEWEB)

Maydanik, Yu. [Ural Branch of the Russian Academy of Sciences, Ekaterinburg (Russian Federation). Inst. of Thermophysics

2004-07-01

Loop heat pipes (LHPs) are hermetic heat-transfer devices operating on a closed evaporation-condensation cycle with the use of capillary pressure for pumping the working fluid [1]. In accordance with this, they possess all the main advantages of conventional heat pipes, but, as distinct from the latter, have a considerably higher heat-transfer capacity, especially when operating in the ''antigravity'' regime, when heat is transferred from above downwards. Besides, LHPs possess a higher functional versatility, are adaptable to different operating conditions and provide great scope for various design embodiments. This is achieved at the expense of both the original design of the device and the properties of the wick - a special capillary structure used for the creation of capillary pressure. The LHP schematic diagram is given in Fig. 1. The device contains an evaporator and a condenser - heat exchanger connected by means of smooth-walled pipe-lines with a relatively small diameter intended for separate motion of vapor and liquid. At present loop heat pipes are most extensively employed in thermoregulation systems of spacecrafts. Miniature LHPs are used for cooling electronics and computers. At the same time there exists a considerable potential of using these devices for the recovery of low-grade (waste) heat from different sources, and also in systems of sun heat supply. In the latter case LHPs may serve as an efficient heat-transfer link between a sun collector and a heat accumulator, which has a low thermal resistance and does not consume any additional energy for pumping the working fluid between them. (orig.)

Nuclear waste storage container with metal matrix

Science.gov (United States)

Sump, Kenneth R.

1978-01-01

The invention relates to a storage container for high-level waste having a metal matrix for the high-level waste, thereby providing greater impact strength for the waste container and increasing heat transfer properties.

Nuclear waste storage container with metal matrix

International Nuclear Information System (INIS)

Sump, K.R.

1978-01-01

The invention relates to a storage container for high-level waste having a metal matrix for the high-level waste, thereby providing greater impact strength for the waste container and increasing heat transfer properties

Passive filtration of air egressing from nuclear containment

Energy Technology Data Exchange (ETDEWEB)

Malloy, III, John D

2017-09-26

A nuclear reactor includes a reactor core comprising fissile material disposed in a reactor pressure vessel. A radiological containment contains the nuclear reactor. A containment compartment contains the radiological containment. A heat sink includes a chimney configured to develop an upward-flowing draft in response to heated fluid flowing into a lower portion of the chimney. A fluid conduit is arranged to receive fluid from the containment compartment and to discharge into the chimney. A filter may be provided, with the fluid conduit including a first fluid conduit arranged to receive fluid from the containment compartment and to discharge into an inlet of the filter, and a second fluid conduit arranged to receive fluid from an outlet of the filter and to discharge into the chimney. As the draft is developed passively, there is no need for a blower or pump configured to move fluid through the fluid conduit.

Thermal convection around a heat source embedded in a box containing a saturated porous medium

Energy Technology Data Exchange (ETDEWEB)

Himasekhar, K.; Bau, H.H. (Univ. of Pennsylvania, Philadelphia (USA))

1988-08-01

A study of the thermal convection around a uniform flux cylinder embedded in a box containing a saturated porous medium is carried out experimentally and theoretically. The experimental work includes heat transfer and temperature field measurements. It is observed that for low Rayleigh numbers, the flow is two dimensional and time independent. Once a critical Rayleigh number is exceeded, the flow undergoes a Hopf bifurcation and becomes three dimensional and time dependent. The theoretical study involves the numerical solution of the two-dimensional Darcy-Oberbeck-Boussinesq equations. The complicated geometry is conveniently handled by mapping the physical domain onto a rectangle via the use of boundary-fitted coordinates. The numerical code can easily be extended to handle diverse geometric configurations. For low Rayleigh numbers, the theoretical results agree favorably with the experimental observations. However, the appearance of three-dimensional flow phenomena limits the range of utility of the numerical code.

Ocean disposal of heat generating radioactive waste

International Nuclear Information System (INIS)

1986-03-01

The objective of this study was to predict tensile stress levels in thin-walled titanium alloy and thick-walled carbon steel containers designed for the ocean disposal of heat-generating radioactive wastes. Results showed that tensile stresses would be produced in both designs by the expansion of the lead filter, for a temperature rise of 200 0 C. Tensile stress could be reduced if the waste heat output at disposal was reduced. Initial stresses for the titanium-alloy containers could be relieved by heat treatment. (UK)

Analysis of large scale tests for AP-600 passive containment cooling system

International Nuclear Information System (INIS)

Sha, W.T.; Chien, T.H.; Sun, J.G.; Chao, B.T.

1997-01-01

One unique feature of the AP-600 is its passive containment cooling system (PCCS), which is designed to maintain containment pressure below the design limit for 72 hours without action by the reactor operator. During a design-basis accident, i.e., either a loss-of-coolant or a main steam-line break accident, steam escapes and comes in contact with the much cooler containment vessel wall. Heat is transferred to the inside surface of the steel containment wall by convection and condensation of steam and through the containment steel wall by conduction. Heat is then transferred from the outside of the containment surface by heating and evaporation of a thin liquid film that is formed by applying water at the top of the containment vessel dome. Air in the annual space is heated by both convection and injection of steam from the evaporating liquid film. The heated air and vapor rise as a result of natural circulation and exit the shield building through the outlets above the containment shell. All of the analytical models that are developed for and used in the COMMIX-ID code for predicting performance of the PCCS will be described. These models cover governing conservation equations for multicomponents single phase flow, transport equations for the κ-ε two-equation turbulence model, auxiliary equations, liquid-film tracking model for both inside (condensate) and outside (evaporating liquid film) surfaces of the containment vessel wall, thermal coupling between flow domains inside and outside the containment vessel, and heat and mass transfer models. Various key parameters of the COMMIX-ID results and corresponding AP-600 PCCS experimental data are compared and the agreement is good. Significant findings from this study are summarized

On-line fouling monitor for heat exchangers

International Nuclear Information System (INIS)

Tsou, J.L.

1995-01-01

Biological and/or chemical fouling in utility service water system heat exchangers adversely affects operation and maintenance costs, and reduced heat transfer capability can force a power deaerating or even a plant shut down. In addition, service water heat exchanger performance is a safety issue for nuclear power plants, and the issue was highlighted by NRC in Generic Letter 89-13. Heat transfer losses due to fouling are difficult to measure and, usually, quantitative assessment of the impact of fouling is impossible. Plant operators typically measure inlet and outlet water temperatures and flow rates and then perform complex calculations for heat exchanger fouling resistance or ''cleanliness''. These direct estimates are often imprecise due to inadequate instrumentation. Electric Power Research Institute developed and patented an on-line condenser fouling monitor. This monitor may be installed in any location within the condenser; does not interfere with routine plant operations, including on-line mechanical and chemical treatment methods; and provides continuous, real-time readings of the heat transfer efficiency of the instrumented tube. This instrument can be modified to perform on-line monitoring of service water heat exchangers. This paper discusses the design, construction of the new monitor, and algorithm used to calculate service water heat exchanger fouling

Container for irradiated fuel

International Nuclear Information System (INIS)

Guy, R.

1978-01-01

The transport container for irradiated or used nuclear fuel is provided with an identical heat shield against fires on the top and bottom sides. Each heat shield consists of two inner nickel plates, whose contact surfaces are polished to a mirror finish and an outer plate of stainless steel. The nickel plate on the box is spot welded to it while the second nickel plate is spot welded to the steel plate. Both together are in turn welded so as to be leaktight to the edges of the box. For extreme heat effects and based on the different (bimetal) coefficients of expansion, the steel plate with the nickel plate attached to it bulges away from the box. The second nickel plate remains at the box, so that a subpressure space is formed with the mirror nickel surfaces. The heat radiation and heat conduction to the box are greatly reduced by this. (DG) [de

Heat treatment of nickel alloys

International Nuclear Information System (INIS)

Smith, D.F. Jr.; Clatworthy, E.F.

1975-01-01

A heat treating process is described that can be used to produce desired combinations of strength, ductility, and fabricability characteristics in heat resistant age-hardenable alloys having precipitation-hardening amounts of niobium, titanium, and/or tantalum in a nickel-containing matrix. (U.S.)

21 CFR 864.9750 - Heat-sealing device.

Science.gov (United States)

2010-04-01

... and Blood Products § 864.9750 Heat-sealing device. (a) Identification. A heat-sealing device is a device intended for medical purposes that uses heat to seal plastic bags containing blood or blood... 21 Food and Drugs 8 2010-04-01 2010-04-01 false Heat-sealing device. 864.9750 Section 864.9750...

Dynamic simulation of periodic adsorption heat pumps. Dynamische Simulation periodischer Adsorptionswaermepumpen

Energy Technology Data Exchange (ETDEWEB)

Foellinger, T.

1989-01-01

Periodic asorption heat pumps with water as working fluid and two types of zeolites as adsorption agents were studied theoretically by a dynamic simulation analysis in order to find out whether they are suited as high-temperature heat pumps for heat recovery. Variants with one and two pairs of containers were investigated. Internal heat transfer is possible between the containers of each pair, and shifting temperature and load profiles (zoned sorption) are generated inside the containers in order to raise the heat ratio (efficience). The heat ratios are clearly higher than in ammonia/water heat pumps of the same size. The external heat transfer is kept constant by means of control elements and buffer systems, so that the periodic heat pump can be integrated in a continuous process. A pilot plant was developed on the basis of the results, with particular interest taken in the design of the liquid/solid heat transfer media. (orig.) With 47 figs., 3 tabs.

CO2 contain of the electric heating

International Nuclear Information System (INIS)

Bacher, P.

2008-02-01

A recent announcement of the RTE and the ADEME on the CO 2 contain of the electric kW, refuting a 2005 study of EDF and ADEME, perturbed the public opinion and was presented as the proof that the nuclear has no part in the fight against the climatic change. The author aims to set things straight. (A.L.B.)

A research on verification of the models in the CONTAIN Code and the uncertainty reduction method for containment integrity evaluation

Energy Technology Data Exchange (ETDEWEB)

Kim, Moo Hwan; Seo, Kyoung Woo [Pohang University of Science and Technology, Pohang (Korea, Republic of)

2000-03-15

The final goal of this research is to evaluate synthetic results of DCH issue and expose accurate methodology to assess containment integrity about operating PWR in Korea. This research is aimed to expose methodology for synthetic resolution of the DCH issue for KSNPP, and make the guide of DCH issue for containment integrity which will be used to design to nuclear power plants.

Removal of sulphur-containing odorants from fuel gases for fuel cell-based combined heat and power applications

Energy Technology Data Exchange (ETDEWEB)

De Wild, P.J.; Nyqvist, R.G.; De Bruijn, F.A.; Stobbe, E.R. [ECN Hydrogen and Clean Fossil Fuels, Petten (Netherlands)

2006-02-15

Natural gas (NG) and liquefied petroleum gas (LPG) are important potential feedstocks for the production of hydrogen for fuel cell-based (e.g. proton exchange membrane fuel cells (PEMFC)) or solid oxide fuel Cells (SOFC) combined heat and power (CHP) applications. To prevent detrimental effects on the (electro)catalysts in fuel cell-based combined heat and power installations (FC-CHP), sulphur removal from the feedstock is mandatory. An experimental bench-marking study of adsorbents has identified several candidates for the removal of sulphur containing odorants at low temperature. Among these adsorbents a new material has been discovered that offers an economically attractive means to remove TetraHydroThiophene (THT), the main European odorant, from natural gas at ambient temperature. The material is environmentally benign, easy to use and possesses good activity (residual sulphur levels below 20 ppbv) and capacity for the common odorant THT in natural gas. When compared to state-of-the-art metal-promoted active carbon the new material has a THT uptake capacity that is up to 10 times larger, depending on temperature and pressure. Promoted versions of the new material have shown potential for the removal of THT at higher temperatures and/or for the removal of other odorants such as mercaptans from natural gas or from LPG.

Removal of sulphur-containing odorants from fuel gases for fuel cell-based combined heat and power applications

Energy Technology Data Exchange (ETDEWEB)

de Wild, P.J.; Nyqvist, R.G.; de Bruijn, F.A.; Stobbe, E.R. [Energy Research Centre of The Netherlands ECN, P.O. Box 1, 1755 ZG Petten (Netherlands)

2006-09-22

Natural gas (NG) and liquefied petroleum gas (LPG) are important potential feedstocks for the production of hydrogen for fuel cell-based (e.g. proton exchange membrane fuel cells (PEMFC) or solid oxide fuel Cells (SOFC) combined heat and power (CHP) applications. To prevent detrimental effects on the (electro)catalysts in fuel cell-based combined heat and power installations (FC-CHP), sulphur removal from the feedstock is mandatory. An experimental bench-marking study of adsorbents has identified several candidates for the removal of sulphur containing odorants at low temperature. Among these adsorbents a new material has been discovered that offers an economically attractive means to remove TetraHydroThiophene (THT), the main European odorant, from natural gas at ambient temperature. The material is environmentally benign, easy to use and possesses good activity (residual sulphur levels below 20ppbv) and capacity for the common odorant THT in natural gas. When compared to state-of-the-art metal-promoted active carbon the new material has a THT uptake capacity that is up to 10 times larger, depending on temperature and pressure. Promoted versions of the new material have shown potential for the removal of THT at higher temperatures and/or for the removal of other odorants such as mercaptans from natural gas or from LPG. (author)

CASTOR {sup registered} HAW28M - a high heat load cask for transport and storage of vitrified high level waste containers

Energy Technology Data Exchange (ETDEWEB)

Vossnacke, A.; Klein, K.; Kuehne, B. [GNS Gesellschaft fuer Nuklear-Service mbH/GNB, Essen (Germany)

2004-07-01

Within the German return programme for vitrified high level waste (HLW) from reprocessing at COGEMA and BNFL up to now 39 casks loaded with 28 containers each were transported back to Germany and are stored in the Interim Storage Facility Gorleben (TBL-G) for up to 40 years. For transport and storage in all but one case the GNB casks CASTOR {sup registered} HAW 20/28 CG have been used. This cask type is designed to accommodate 20 or 28 HLW containers with a total thermal power of 45 kW maximum. In the near future, among the high level waste, which has to be returned to Germany, there will be an increasing number of containers of which the heat capacity and radioactive inventory will exceed the technical limits of the CASTOR {sup registered} HAW 20/28 CG. Therefore GNB has started the development of a new cask generation, named CASTOR {sup registered} HAW28M, meeting these future requirements. The CASTOR {sup registered} HAW28M is especially developed for the transport of vitrified residues from France and Great Britain to Germany. It complies with the international regulations for type B packages according to IAEA (International Atomic Energy Agency). It is thus guaranteed that even in case of any accident the cask body and the lid system remain functional and the safe confinement of the radioactive contents remains intact during transport. The CASTOR {sup registered} HAW28M fulfills not only the requirements for transport but also the acceptance criteria of interim storage: radiation shielding, heat dissipation, safe confinement under both normal and hypothetical accident conditions. Storage buildings such as the TBL-G simply support the safety functions of the cask. The challenge for the development results from higher requirements of the technical specification, particularly related to fuel which is reprocessed. As a consequence of the reprocessing of fuel with increased enrichment and burn up, higher heat capacity and sophisticated shielding measures have to be

16 CFR 1500.130 - Self-pressurized containers: labeling.

Science.gov (United States)

2010-01-01

... 16 Commercial Practices 2 2010-01-01 2010-01-01 false Self-pressurized containers: labeling. 1500... § 1500.130 Self-pressurized containers: labeling. (a) Self-pressurized containers that fail to bear a...: warning—contents under pressure Do not puncture or incinerate container. Do not expose to heat or store at...

Sensitivity of Sump Water Temperature to Containment Integrity

Energy Technology Data Exchange (ETDEWEB)

Jang, Misuk; Kim, Seoung Rae [Nuclear Engineering Service and Solution, Daejeon (Korea, Republic of)

2014-05-15

This paper is focused on the containment behavior analysis in the above described cases using GOTHIC-IST (generation of thermal-hydraulic information for containments, industry standard toolset). GOTHIC-IST version 7.2a is an integrated, general purpose thermal-hydraulics software package for design, licensing, safety and operating analysis of nuclear power plant containments and other confinement buildings. In this study, we perform the sensitivity the sump water temperature to containment integrity. For 35% RIH break accident with the malfunction of spray system, local air coolers, ECC(emergency core cooling) pump and heat exchanger, the peak pressure at boiler room do not exceed the design pressure 124kPa(g) of the containment and containment integrity is secured. If accompanied the malfunction of heat exchanger or pump in the time of low pressure safety injection, of ECCS, it will be one of the aggravating factors to the integrity of core and containment.

Stress corrosion cracking of steam generator tubing materials in lead containing solution

International Nuclear Information System (INIS)

Kim, H.P.; Hwang, S.S.; Kim, J.S.; Hong, J.H.

2007-01-01

Stress corrosion cracking (SCC) in lead (Pb) containing environments has been one of key issues in the nuclear power industry since Pb had been identified as a cause of the SCC of steam generator (SG) tubing materials in some power plants. To mitigate or prevent degradation of SG tubing materials, a mechanistic understanding of SCC in Pb containing environment is needed, along with an understanding of the source and transport behaviors of Pb species in the secondary circuit. In this work, SCC behaviors of Alloy 600 in Pb containing environments were studied. Influences of microstructures of Alloy 600 and the inhibitive additives were investigated using the C-ring and the slow strain rate tests in caustic solution and demineralized water at 315 o C. Microstructures of Alloy 600 were varied by heat treatment at different temperatures. The additives examined were nickel boride (NiB) and cerium boride (CeB 6 ). The surface films were analyzed using Auger Electron Spectroscopy (AES) and Energy Dispersive X-ray Spectroscopy (EDS). The SCC mode varied with microstructure. Effectiveness of the additives in Pb containing environments is discussed. (author)

Three dimensional modeling on airflow, heat and mass transfer in partially impermeable enclosure containing agricultural produce during natural convective cooling

International Nuclear Information System (INIS)

Chourasia, M.K.; Goswami, T.K.

2007-01-01

A three dimensional model was developed to simulate the transport phenomena in heat and mass generating porous medium cooled under natural convective environment. Unlike the previous works on this aspect, the present model was aimed for bulk stored agricultural produce contained in a permeable package placed on a hard surface. This situation made the bottom of the package impermeable to fluid flow as well as moisture transfer and adiabatic to heat transfer. The velocity vectors, isotherms and contours of rate of moisture loss were presented during transient cooling as well as at steady state using the commercially available computational fluid dynamics (CFD) code based on the finite volume technique. The CFD model was validated using the experimental data on the time-temperature history as well as weight loss obtained from a bag of potatoes kept in a cold store. The simulated and experimental values on temperature and moisture loss of the product were found to be in good agreement

Liquid metal reactor applications of the CONTAIN code

International Nuclear Information System (INIS)

Carroll, D.E.; Bergeron, K.D.; Gido, R.; Valdez, G.D.; Scholtyssek, W.

1988-01-01

The CONTAIN code is the NRC's best-estimate code for the evaluation of the conditions that may exist inside a reactor containment building during a severe accident. Included in the phenomena modeled are thermal-hydraulics, radiant and convective heat transfer, aerosol loading and transient response, fission product transport and heating effects, and interactions of sodium and corium with the containment atmosphere and structures. CONTAIN has been used by groups in Japan and West Germany to assess its ability to analyze accident consequences for liquid metal reactor (LMR) plants. In conjunction with this use, collaborative efforts to improve the modeling have been pursued. This paper summarizes the current state of the version of CONTAIN that has been enhanced with extra capabilities for LMR applications. A description of physical models is presented, followed by a review of validation exercises performed with CONTAIN. Some demonstration calculations of an integrated LMR application are presented

The use of ferrofluids for heat removal: Advantage or disadvantage?

Energy Technology Data Exchange (ETDEWEB)

Krauzina, Marina T., E-mail: krauzina@psu.ru [Faculty of Physics, Perm State University, 15 Bukirev Street, Perm 614990 (Russian Federation); Bozhko, Aleksandra A., E-mail: bozhko@psu.ru [Faculty of Physics, Perm State University, 15 Bukirev Street, Perm 614990 (Russian Federation); Krauzin, Pavel V., E-mail: krauzin@psu.ru [Faculty of Physics, Perm State University, 15 Bukirev Street, Perm 614990 (Russian Federation); Suslov, Sergey A., E-mail: ssuslov@swin.edu.au [Department of Mathematics H38, Swinburne University of Technology, Hawthorn, Victoria 3122 (Australia)

2017-06-01

It is shown experimentally that, depending on the relative orientation of the gravity and the thermal gradient and on the pre-history of experiment, the application of a uniform external vertical magnetic field to a spherical cavity filled with magnetic ferrofluid can either enhance or suppress a convective heat transfer. - Highlights: • Conduction heat transfer in magnetic fluid heated from above is stronger than that in a fluid not containing nanoparticles. • The application of a uniform vertical magnetic field enhances heat transfer when magnetic fluid is heated from above. • Heat transfer in a magnetic fluid heated from below is weaker than that in a fluid not containing nanoparticles.

Heat pipes. (Latest citations from the NTIS bibliographic database). Published Search

Energy Technology Data Exchange (ETDEWEB)

NONE

1995-10-01

The bibliography contains citations concerning the theory, design, fabrication, testing, and operation of heat pipes. Applications include heat rejection devices in spacecraft, use in passive solar heating systems and warm air furnaces, and electronic circuit cooling. Heat recovery operations, and materials considerations are also discussed.(Contains 50-250 citations and includes a subject term index and title list.) (Copyright NERAC, Inc. 1995)

Light and heavy water replacing system in reactor container

International Nuclear Information System (INIS)

Miyamoto, Keiji.

1979-01-01

Purpose: To enable to determine the strength of a reactor container while neglecting the outer atmospheric pressure upon evacuation, by evacuating the gap between the reactor container and a biological thermal shield, as well as the container simultaneously upon light water - heavy water replacement. Method: Upon replacing light water with heavy water by vacuum evaporation system in a nuclear reactor having a biological thermal shield surrounding the reactor container incorporating therein a reactor core by way of a heat expansion absorbing gap, the reactor container and the havy water recycling system, as well as the inside of heat expansion absorbing gap are evacuated simultaneously. This enables to neglect the outer atmospheric outer pressure upon evacuation in the determination of the container strength, and the thickness of the container can be decreased by so much as the external pressure neglected. (Moriyama, K.)

Cutaneous C-polymodal fibers lacking TRPV1 are sensitized to heat following inflammation, but fail to drive heat hyperalgesia in the absence of TPV1 containing C-heat fibers

Directory of Open Access Journals (Sweden)

Koerber H Richard

2010-09-01

Full Text Available Abstract Background Previous studies have shown that the TRPV1 ion channel plays a critical role in the development of heat hyperalgesia after inflammation, as inflamed TRPV1-/- mice develop mechanical allodynia but fail to develop thermal hyperalgesia. In order to further investigate the role of TRPV1, we have used an ex vivo skin/nerve/DRG preparation to examine the effects of CFA-induced-inflammation on the response properties of TRPV1-positive and TRPV1-negative cutaneous nociceptors. Results In wildtype mice we found that polymodal C-fibers (CPMs lacking TRPV1 were sensitized to heat within a day after CFA injection. This sensitization included both a drop in average heat threshold and an increase in firing rate to a heat ramp applied to the skin. No changes were observed in the mechanical response properties of these cells. Conversely, TRPV1-positive mechanically insensitive, heat sensitive fibers (CHs were not sensitized following inflammation. However, results suggested that some of these fibers may have gained mechanical sensitivity and that some previous silent fibers gained heat sensitivity. In mice lacking TRPV1, inflammation only decreased heat threshold of CPMs but did not sensitize their responses to the heat ramp. No CH-fibers could be identified in naïve nor inflamed TRPV1-/- mice. Conclusions Results obtained here suggest that increased heat sensitivity in TRPV1-negative CPM fibers alone following inflammation is insufficient for the induction of heat hyperalgesia. On the other hand, TRPV1-positive CH fibers appear to play an essential role in this process that may include both afferent and efferent functions.

TENERIFE program: high temperature experiments on A 4 tons UF6 container

International Nuclear Information System (INIS)

Casselman, C.; Duret, B.; Seiler, J.M.; Ringot, C.; Warniez, P.; Wataru, M.; Shiomi, S.; Ozaki, S.; Yamakawa, H.

1993-01-01

To know the input of the future thermo-mechanical code, we have to get a better understanding of the thermo-physical evolution of the UF 6 which pressurizes the container. This evolution is function of: a) the heat transfer rate from the fire to the container b) the UF 6 behaviour in the container. These tests are essentially analytical at simulated fire temperatures of between 800 and 1000degC. They use a representative mass of UF 6 (around 4 tons). The tests will not seek to rupture the test container which has a diameter equal to the 48Y container, but shorter length. These tests carried out in realistic conditions (typical thermal gradient at the wall, characteristic period for UF 6 internal mass transfer) should make possible to improve knowledge of two fundamental phenomena: 1) vaporization of UF 6 on contact with the heated wall (around 400degC), a phenomenon which controls the container internal pressurization kinetic, 2) the equivalent conductivity of solid UF 6 , a phenomenon which is linked to the heat transfer by UF 6 vaporization-condensation through the solid's porosities and which depends on the diameter of the container. In addition, they will allow the influence of other parameters to be studied, such as UF 6 container filling mode or the mechanical characteristics of the container material. A UF 6 container fitted with instruments (wall temperature, UF 6 temperature, pressure) is heated by a rapid heat transient in a radiating furnace where the temperature and thermal power supplied can be measured. The test continues until pre-established thresholds have been reached: 1) strain threshold measured on the container surface (strain gauges positioned on the outside), 2) maximum temperature threshold of UF 6 , 3) container internal pressure threshold. (J.P.N.)

A passive emergency heat sink for water-cooled reactors with particular application to CANDU reactors

International Nuclear Information System (INIS)

Spinks, N.J.

1996-01-01

Water in an overhead pool can serve as a general-purpose passive emergency heat sink for water-cooled reactors. It can be used for containment cooling, for emergency depressurization of the heat transport-system, or to receive any other emergency heat, such as that from the CANDU moderator. The passive emergency water system provides in-containment depressurization of steam generators and no other provision is needed for supply of low-pressure emergency water to the steam generators. For containment cooling, the pool supplies water to the tube side of elevated tube banks inside containment. The elevation with respect to the reactor heat source maximizes heat transport, by natural convection, of hot containment gases. This effective heat transport combines with the large heat-transfer coefficients of tube banks, to reduce containment overpressure during accidents. Cooled air from the tube banks is directed past the break in the heat-transport system, to facilitate removal of hydrogen using passive catalytic recombiners. (author)

Effect of yttrium on the oxide scale adherence of pre-oxidized silicon-containing heat-resistant alloy

International Nuclear Information System (INIS)

Yan Jingbo; Gao Yimin; Shen Yudi; Yang Fang; Yi Dawei; Ye Zhaozhong; Liang Long; Du Yingqian

2011-01-01

Highlights: → AE experiment shows yttrium has a beneficial effect on the pre-oxidized HP40 alloy. → Yttrium facilitates the formation of internal oxide after 10 h of oxidation. → Internal oxide changes the rupture behaviour of the oxide scale. → Twins form in the internal oxide and improve the binding strength of the scale. - Abstract: This paper investigates the effect of the rare earth element yttrium on the rupture behaviour of the oxide scale on the silicon-containing heat-resistant alloy during cooling. After 10 h of oxidation, yttrium is found to facilitate the formation of internal oxides (silica) at the scale-matrix interface. Due to the twinning observed by scanning transmission electron microscopy (STEM) in silica, the critical strain value for the scale failure can be dramatically improved, and the formation of cracks at the scale-matrix interface is inhibited.

CONTEMPT, LWR Containment Pressure and Temperature Distribution in LOCA

International Nuclear Information System (INIS)

Hargroves, D.W.; Metcalfe, L.J.; Cheng, Teh-Chin; Wheat, L.L.; Mings, W.J.

1991-01-01

1 - Description of problem or function: CONTEMPT-LT was developed to predict the long-term behavior of water-cooled nuclear reactor containment systems subjected to postulated loss-of-coolant accident (LOCA) conditions. CONTEMPT-LT calculates the time variation of compartment pressures, temperatures, mass and energy inventories, heat structure temperature distributions, and energy exchange with adjacent compartments. The program is capable of describing the effects of leakage on containment response. Models are provided for fan cooler and cooling spray engineered safety systems. One to four compartments can be modeled, and any compartment except the reactor system may have both a liquid pool region and an air-vapor atmosphere region above the pool. Each region is assumed to have a uniform temperature, but the temperatures of the two regions may be different. The user determines the compartments to be used, specifies input mass and energy additions, defines heat structure and leakage systems, and prescribes the time advancement and output control. CONTEMPT-LT/28-H (NESC0433/08) includes also models for hydrogen combustion. 2 - Method of solution: The initial conditions of the containment atmosphere are calculated from input values, and the initial temperature distributions through the containment structures are determined from the steady-state solution of the heat conduction equations. A time advancement proceeds as follows. The input water and energy rates are evaluated at the midpoint of a time interval and added to the containment system. Pressure suppression, spray system effects, and fan cooler effects are calculated using conditions at the beginning of a time-step. Leakage and heat losses or gains, extrapolated from the last time-step, are added to the containment system. Containment volume pressure and temperature are estimated by solving the mass, volume, and energy balance equations. Using these results as boundary conditions, the heat conduction equations

Dynamic Complexity Study of Nuclear Reactor and Process Heat Application Integration

Energy Technology Data Exchange (ETDEWEB)

J' Tia Patrice Taylor; David E. Shropshire

2009-09-01

Abstract This paper describes the key obstacles and challenges facing the integration of nuclear reactors with process heat applications as they relate to dynamic issues. The paper also presents capabilities of current modeling and analysis tools available to investigate these issues. A pragmatic approach to an analysis is developed with the ultimate objective of improving the viability of nuclear energy as a heat source for process industries. The extension of nuclear energy to process heat industries would improve energy security and aid in reduction of carbon emissions by reducing demands for foreign derived fossil fuels. The paper begins with an overview of nuclear reactors and process application for potential use in an integrated system. Reactors are evaluated against specific characteristics that determine their compatibility with process applications such as heat outlet temperature. The reactor system categories include light water, heavy water, small to medium, near term high-temperature, and far term high temperature reactors. Low temperature process systems include desalination, district heating, and tar sands and shale oil recovery. High temperature processes that support hydrogen production include steam reforming, steam cracking, hydrogen production by electrolysis, and far-term applications such as the sulfur iodine chemical process and high-temperature electrolysis. A simple static matching between complementary systems is performed; however, to gain a true appreciation for system integration complexity, time dependent dynamic analysis is required. The paper identifies critical issues arising from dynamic complexity associated with integration of systems. Operational issues include scheduling conflicts and resource allocation for heat and electricity. Additionally, economic and safety considerations that could impact the successful integration of these systems are considered. Economic issues include the cost differential arising due to an integrated

Guidelines for District Heating Substations

Energy Technology Data Exchange (ETDEWEB)

2009-07-15

The present guidelines contain a set of recommendations focusing on planning, installation, use and maintenance of district heating (DH) substations within district heating systems throughout Europe. The recommendations were developed in order to enable readers to develop well-functioning substations and an effective heat and domestic warm water delivery. These guidelines are intended to give the most effective overall solutions for various parts of the customer installation. The guidelines are not meant to specify the different components of the substation such as meters or heat exchangers. The guidelines deal with a wide variety of issues concerning both present systems of today and district heating systems of the future. Specific handling and maintenance recommendations are mainly focused on present modern systems but are also intended to cover the future situation as much as is feasible. For this reason, certain existing systems are not dealt with in these guidelines. For instance, these guidelines do not cover steam systems, systems with temperatures exceeding 110 deg C and pressure levels above 1.6 MPa. The guidelines include a chapter on the heat meter, as the meter and especially the meter installation is always installed simultaneously with the rest of the substation. These guidelines aim to provide best-practice and easy-to-handle recommendations for: - those who are responsible for relations between district heating utilities and customers; - those who own or maintain a building connected to the district heating network; - those who manufacture, plan, purchase, test and install substations. These guidelines do not deal with investment or cost aspects, but in general, Euroheat and Power recommends looking at the lifetime cost of all components of the substation, instead of investment costs alone. An example of this is provided in Chapter 7.8. The Guidelines were developed based on the most optimal operating principles of substations and meters

Concrete and prestressing process, container made with this concrete

International Nuclear Information System (INIS)

Gerard, M.

1992-01-01

Shape memory alloy fibers or heat shrinking fibers are encapsulated in a standard concrete. Prestressed concrete is obtained by heat treatment. Application is made to the fabrication of radioactive waste containers

Closed loop solar chemical heat pipe

International Nuclear Information System (INIS)

Levy, M.; Levitan, R.; Rosin, H.; Rubin, R.

1991-01-01

The system used for the closed loop operation of the solar chemical heat pipe comprises a reformer, heated by the solar furnace, a methanator and a storage assembly containing a compressor and storage cylinders. (authors). 7 figs

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.

District heating versus local heating - Social supportability

International Nuclear Information System (INIS)

Matei, Magdalena; Enescu, Diana; Varjoghie, Elena; Radu, Florin; Matei, Lucian

2004-01-01

District heating, DH, is an energy source which can provide a cost-effective, environmentally friendly source of heat and power for cities, but only in the case of well running systems, with reasonable technological losses. The benefits of DH system are well known: environmental friendly, energy security, economic and social advantages. DH already covers 60% of heating and hot water needs in transition economies. Today, 70 % of Russian, Latvian and Belarus homes use DH, and heating accounts for one-third of total Russian energy consumption. Yet a large number of DH systems in the region face serious financial, marketing or technical problems because of the policy framework. How can DH issues be best addressed in national and local policy? What can governments do to create the right conditions for the sustainable development of DH while improving service quality? What policies can help capture the economic, environmental and energy security benefits of co-generation and DH? To address these questions, the International Energy Agency (IEA) hosted in 2002 and 2004 conference focusing on the crucial importance of well-designed DH policies, for exchanging information on policy approaches. The conclusions of the conference have shown that 'DH systems can do much to save energy and boost energy security, but stronger policy measures are needed to encourage wise management and investment. With a stronger policy framework, DH systems in formerly socialist countries could save the equivalent of 80 billion cubic meters of natural gas a year through supply side efficiency improvements. This is greater than total annual natural gas consumption in Italy'. More efficient systems will also decrease costs, reducing household bills and making DH competitive on long-term. This paper presents the issues: -Theoretical benefits of the district heating and cooling systems; - Municipal heating in Romania; - Technical and economic problems of DH systems and social supportability; - How

Control challenges in domestic heating systems

DEFF Research Database (Denmark)

Thybo, Honglian; Larsen, Lars F. S.; Weitzmann, Peter

2007-01-01

The objective of this paper is to analyze domestic heating applications and identify unfavorable building constructions and control challenges to be addressed by high performance heating control systems. Heating of domestic houses use a large amount of the total energy consumption in Scandinavia....... Hence the potential of reducing energy consumption by applying high performance control is vast. Indoor climate issues are becoming more in focus, which also leads to a demand for high performance heating systems. The paper presents an analysis of how the building elements of today's domestic houses...... with water based floor heating affect the control challenge. The analysis is documented with simulation results....

1995 national heat transfer conference: Proceedings. Volume 12: Falling films; Fundamentals of subcooled flow boiling; Compact heat exchanger technology for the process industry; HTD-Volume 314

International Nuclear Information System (INIS)

Sernas, V.; Boyd, R.D.; Jensen, M.K.

1995-01-01

The papers in the first section cover falling films and heat transfer. Papers in the second section address issues associated with heat exchangers, such as: plate-and-frame heat exchanger technology; thermal design issues; condensation; and single-phase flows. The papers in the third section deal with studies related to: the turbulent velocity field in a vertical annulus; the effects of curvature and a dissolved noncondensable gas on nucleate boiling heat transfer; the effects of flow obstruction on the onset of a Ledinegg-type flow instability; pool boiling from a large-diameter tube; and two-dimensional wall temperature distributions and convection in a single-sided heated vertical tube. Separate abstracts were prepared for most papers in this volume

Cavity pressure history of contained nuclear explosions

Energy Technology Data Exchange (ETDEWEB)

Chapin, C E [Lawrence Radiation Laboratory, University of California, Livermore, CA (United States)

1970-05-01

Knowledge of pressure in cavities created by contained nuclear explosions is useful for estimating the possibility of venting radioactive debris to the atmosphere. Measurements of cavity pressure, or temperature, would be helpful in evaluating the correctness of present code predictions of underground explosions. In instrumenting and interpreting such measurements it is necessary to have good theoretical estimates of cavity pressures. In this paper cavity pressure is estimated at the time when cavity growth is complete. Its subsequent decrease due to heat loss from the cavity to the surrounding media is also predicted. The starting pressure (the pressure at the end of cavity growth) is obtained by adiabatic expansion to the final cavity size of the vaporized rock gas sphere created by the explosion. Estimates of cavity size can be obtained by stress propagation computer codes, such as SOC and TENSOR. However, such estimates require considerable time and effort. In this paper, cavity size is estimated using a scheme involving simple hand calculations. The prediction is complicated by uncertainties in the knowledge of silica water system chemistry and a lack of information concerning possible blowoff of wall material during cavity growth. If wall material blows off, it can significantly change the water content in the cavity, compared to the water content in the ambient media. After cavity growth is complete, the pressure will change because of heat loss to the surrounding media. Heat transfer by convection, radiation and conduction is considered, and its effect on the pressure is calculated. Analysis of cavity heat transfer is made difficult by the complex nature of processes which occur at the wall where melting, vaporization and condensation of the gaseous rock can all occur. Furthermore, the melted wall material could be removed by flowing or dripping to the cavity floor. It could also be removed by expansion of the steam contained in the melt (blowoff) and by

Process for introducing radioactive articles into a transport and/or storage container and transporting and/or storing the container and later extraction of the article from the container, and container for transporting and/or storing radioactive articles

International Nuclear Information System (INIS)

Vox, A.J.

1979-01-01

The articles, for example fuel elements, are introduced into the container and the remaining space inside the container is filled with lead, a salt or a mixture of salts of eutectic composition, which freezes at ambient temperature. This makes dry transport possible. To extract the fuel elements, it is sufficient to heat the container, which softens the protective and shielding material. The salt or mixture of salts is suitable for thermal conduction. (DG) [de

Improved solar heating systems

Science.gov (United States)

Schreyer, J.M.; Dorsey, G.F.

1980-05-16

An improved solar heating system is described in which the incident radiation of the sun is absorbed on collector panels, transferred to a storage unit and then distributed as heat for a building and the like. The improvement is obtained by utilizing a storage unit comprising separate compartments containing an array of materials having different melting points ranging from 75 to 180/sup 0/F. The materials in the storage system are melted in accordance with the amount of heat absorbed from the sun and then transferred to the storage system. An efficient low volume storage system is provided by utilizing the latent heat of fusion of the materials as they change states in storing ad releasing heat for distribution.

Condensation in the presence of noncondensible gases: AP600 containment simulation

Energy Technology Data Exchange (ETDEWEB)

Anderson, M.H.; Corradini, M.L.

1995-09-01

The Westinghouse Electric Corporation has designed an advanced pressurized light water reactor, AP600. This reactor is designed with a passive cooling system to remove sensible and decay heat from the containment. The heat removal path involves condensation heat transfer, aided by natural convective forces generated by buoyancy effects. A one-twelfth scale rectangular slice of the proposed reactor containment was constructed at the University of Wisconsin to simulate conditions anticipated from transients and accidents that may occur in a full scale containment vessel under a variety of conditions. Similitude of the test facility was obtained by considering the appropriate dimensionless group for the natural convective process (modified Froude number) and the aspect ratio (H/R) of the containment vessel. An experimental investigation to determine the heat transfer coefficients associated with condensation on a vertical and horizontal cooled wall (located in the scaled test section) at several different inlet steam flow rates and test section temperatures was conducted. In this series of experiments, the non-condensible mass fraction varied between (0.9-0.4) with corresponding mixture temperatures between 60-90{degrees}C. The heat transfer coefficients of the top horizontal surface varied from (82-296)W/m{sup 2}K and the vertical side heat transfer coefficients varied form (70-269)m{sup 2}K. The results were then compared to boundary layer heat and mass transfer theory by the use of the McAdams correlation for free convection.

Explosion testing for the container venting system

International Nuclear Information System (INIS)

Cashdollar, K.L.; Green, G.M.; Thomas, R.A.; Demiter, J.A.

1993-01-01

As part of the study of the hazards of inspecting nuclear waste stored at the Hanford Site, the US Department of Energy and Westinghouse Hanford Company have developed a container venting system to sample the gases that may be present in various metal drums and other containers. In support of this work, the US Bureau of Mines has studied the probability of ignition while drilling into drums and other containers that may contain flammable gas mixtures. The Westinghouse Hanford Company drilling procedure was simulated by tests conducted in the Bureau's 8-liter chamber, using the same type of pneumatic drill that will be used at the Hanford Site. There were no ignitions of near-stoichiometric hydrogen-air or methane-air mixtures during the drilling tests. The temperatures of the drill bits and lids were measured by an infrared video camera during the drilling tests. These measured temperatures are significantly lower than the ∼500 degree C autoignition temperature of uniformly heated hydrogen-air or the ∼600 degree C autoignition temperature of uniformly heated methane-air. The temperatures are substantially lower than the 750 degree C ignition temperature of hydrogen-air and 1,220 degree C temperature of methane-air when heated by a 1-m-diameter wire

Cascade: a review of heat transport and plant design issues

International Nuclear Information System (INIS)

Murray, K.A.; McDowell, M.W.

1984-01-01

A conceptual heat transfer loop for Cascade, a centrifugal-action solid-breeder reaction chamber, has been investigated and results are presented. The Cascade concept, a double-cone-shaped reaction chamber, rotates along its horizontal axis. Solid Li 2 O or other lithium-ceramic granules are injected tangentially through each end of the chamber. The granules cascade axially from the smaller radii at the ends to the larger radius at the center, where they are ejected into a stationary granule catcher. Heat and tritium are then removed from the granules and the granules are reinjected into the chamber. A 50% dense Li 2 O granule throughput of 2.8 m 3 /s is transferred from the reaction chamber to the steam generators via continuous bucket elevators. The granules then fall by gravity through 4 vertical steam generators. The entire transport system is maintained at the same vacuum conditions present inside the reaction chamber

49 CFR 179.220-11 - Postweld heat treatment.

Science.gov (United States)

2010-10-01

... 49 Transportation 2 2010-10-01 2010-10-01 false Postweld heat treatment. 179.220-11 Section 179... Postweld heat treatment. (a) Postweld heat treatment of the inner container is not a specification requirement. (b) Postweld heat treatment of the cylindrical portions of the outer shell to which the anchorage...

Corrosion issues in high-level nuclear waste containers

Science.gov (United States)

Asl, Samin Sharifi

In this dissertation different aspects of corrosion and electrochemistry of copper, candidate canister material in Scandinavian high-level nuclear waste disposal program, including the thermodynamics and kinetics of the reactions that are predicted to occur in the practical system have been studied. A comprehensive thermodynamic study of copper in contact with granitic groundwater of the type and composition that is expected in the Forsmark repository in Sweden has been performed. Our primary objective was to ascertain whether copper would exist in the thermodynamically immune state in the repository, in which case corrosion could not occur and the issue of corrosion in the assessment of the storage technology would be moot. In spite of the fact that metallic copper has been found to exist for geological times in granitic geological formations, copper is well-known to be activated from the immune state to corrode by specific species that may exist in the environment. The principal activator of copper is known to be sulfur in its various forms, including sulfide (H2S, HS-, S2-), polysulfide (H2Sx, HSx -, Sx 2-), poly sulfur thiosulfate ( SxO3 2-), and polythionates (SxO6 2-). A comprehensive study of this aspect of copper chemistry has never been reported, and yet an understanding of this issue is vital for assessing whether copper is a suitable material for fabricating canisters for the disposal of HLNW. Our study identifies and explores those species that activate copper; these species include sulfur-containing entities as well as other, non-sulfur species that may be present in the repository. The effects of temperature, solution pH, and hydrogen pressure on the kinetics of the hydrogen electrode reaction (HER) on copper in borate buffer solution have been studied by means of steady-state polarization measurements, including electrochemical impedance spectroscopy (EIS). In order to obtain electrokinetic parameters, such as the exchange current density and the

Improvements in or relating to microwave heating

International Nuclear Information System (INIS)

Hardwick, W.H.; Gayler, R.

1984-01-01

The invention is concerned with microwave heating and more particularly with microwave heating to effect melting. Material is contained in a container and microwave energy is applied to melt material while a portion of the material adjacent to the internal surfaces of the container is maintained at a temperature at which it is not molten. This may be by cooling water in vessel. The invention therefore provides for ''skull'' melting in which a layer of non-molten material protects the container from molten material. Examples of materials which may be treated in accordance with the present invention are glasses and glass-like materials and precursors therefor (e.g. glasses and glass-like materials or precursors therefor containing radioactive wastes). (author)

Comparative analysis of heat pump and biomass boiler for small detached house heating

Directory of Open Access Journals (Sweden)

Olkowski Tomasz

2017-01-01

Full Text Available The purpose of the work is to answer the question - which of the two selected heat sources is more economically beneficial for small detached house: heat pump or biomass boiler fuelled with wood-pellets? The comparative analysis of these sources was carried out to discuss the issue. First, cost of both, equipment and operation of selected heat systems were analysed. Additionally, CO2 emission levels associated with these heat systems were determined. The comparative analysis of the costs of both considered heat systems showed that equipment cost of heat pump system is considerably bigger than the cost of biomass boiler system. The comparison of annual operation costs showed that heat pump operation cost is slightly lower than operation cost of biomass boiler. The analysis of above results shows that lower operation cost of heat pump in comparison with biomass boiler cost lets qualify heat pump as more economically justified only after 38 years of work. For both analysed devices, CO2 emission levels were determined. The considerations take into account the fact that heat pump consumes electricity. It is mostly generated through combustion of coal in Poland. The results show that in Poland biomass boiler can be described as not only more economically justified system but also as considerably more ecological.

Comparative analysis of heat pump and biomass boiler for small detached house heating

Science.gov (United States)

Olkowski, Tomasz; Lipiński, Seweryn; Olędzka, Aneta

2017-10-01

The purpose of the work is to answer the question - which of the two selected heat sources is more economically beneficial for small detached house: heat pump or biomass boiler fuelled with wood-pellets? The comparative analysis of these sources was carried out to discuss the issue. First, cost of both, equipment and operation of selected heat systems were analysed. Additionally, CO2 emission levels associated with these heat systems were determined. The comparative analysis of the costs of both considered heat systems showed that equipment cost of heat pump system is considerably bigger than the cost of biomass boiler system. The comparison of annual operation costs showed that heat pump operation cost is slightly lower than operation cost of biomass boiler. The analysis of above results shows that lower operation cost of heat pump in comparison with biomass boiler cost lets qualify heat pump as more economically justified only after 38 years of work. For both analysed devices, CO2 emission levels were determined. The considerations take into account the fact that heat pump consumes electricity. It is mostly generated through combustion of coal in Poland. The results show that in Poland biomass boiler can be described as not only more economically justified system but also as considerably more ecological.

Westinghouse-GOTHIC comparisons to AP600 passive containment cooling tests

International Nuclear Information System (INIS)

Kennedy, M.D.; Woodcock, J.; Gresham, J.A.

1994-01-01

Westinghouse-GOTHIC is a thermal-hydraulics code well suited to analyzing passively cooled containments which depend on heat removal primarily through the containment shell. The code includes boundary layer heat and mass transfer correlations. A liquid film convective energy transport model has been added to the Westinghouse-GOTHIC code to account for the sensible heat change of the applied exterior water. The objective of this paper is to compare the code's predictions of the AP600 large scale test facility with and without the liquid film convective energy transport model. The predicted vessel pressure and integrated heat rate with and without the film convective energy transport model will be compared to the measured data. (author)

Containment atmosphere response to external sprays

International Nuclear Information System (INIS)

Green, J.; Almenas, K.

1995-01-01

The application of external sprays to a containment steel shell can be an effective energy removal method and has been proposed in the passive AP-600 design. Reduction of the steel shell temperature in contact with the containment atmosphere enhances both heat and mass transfer driving forces. Large scale experimental data in this area is scarce, therefore the measurements obtained from the E series tests conducted at the German HDR facility deserve special attention. These long term tests simulated various severe accident conditions, including external spraying of the hemispherical steel shell. This investigation focuses upon the integral response of the HDR containment atmosphere during spray periods and upon methods by which lumped parameter system codes, like CONTAIN, model the underlying condensation phenomena. Increases in spray water flowrates above a minimum value were ineffective at improving containment pressure reduction since the limiting resistance for energy transfer lies in the noncondensable-vapor boundary layer at the inner condensing surface. The spray created an unstable condition by cooling the upper layers of a heated atmosphere and thus inducing global natural circulation flows in the facility and subsequently, abrupt changes in lighter-than-air noncondensable (J 2 /He) concentrations. Modeling results using the CONTAIN code are outlined and code limitations are delineated

Containment atmosphere response to external sprays

Energy Technology Data Exchange (ETDEWEB)

Green, J.; Almenas, K. [Univ. of Maryland, College Park, MD (United States)

1995-09-01

The application of external sprays to a containment steel shell can be an effective energy removal method and has been proposed in the passive AP-600 design. Reduction of the steel shell temperature in contact with the containment atmosphere enhances both heat and mass transfer driving forces. Large scale experimental data in this area is scarce, therefore the measurements obtained from the E series tests conducted at the German HDR facility deserve special attention. These long term tests simulated various severe accident conditions, including external spraying of the hemispherical steel shell. This investigation focuses upon the integral response of the HDR containment atmosphere during spray periods and upon methods by which lumped parameter system codes, like CONTAIN, model the underlying condensation phenomena. Increases in spray water flowrates above a minimum value were ineffective at improving containment pressure reduction since the limiting resistance for energy transfer lies in the noncondensable-vapor boundary layer at the inner condensing surface. The spray created an unstable condition by cooling the upper layers of a heated atmosphere and thus inducing global natural circulation flows in the facility and subsequently, abrupt changes in lighter-than-air noncondensable (J{sub 2}/He) concentrations. Modeling results using the CONTAIN code are outlined and code limitations are delineated.

Code manual for CONTAIN 2.0: A computer code for nuclear reactor containment analysis

International Nuclear Information System (INIS)

Murata, K.K.; Williams, D.C.; Griffith, R.O.; Gido, R.G.; Tadios, E.L.; Davis, F.J.; Martinez, G.M.; Washington, K.E.; Tills, J.

1997-12-01

The CONTAIN 2.0 computer code is an integrated analysis tool used for predicting the physical conditions, chemical compositions, and distributions of radiological materials inside a containment building following the release of material from the primary system in a light-water reactor accident. It can also predict the source term to the environment. CONTAIN 2.0 is intended to replace the earlier CONTAIN 1.12, which was released in 1991. The purpose of this Code Manual is to provide full documentation of the features and models in CONTAIN 2.0. Besides complete descriptions of the models, this Code Manual provides a complete description of the input and output from the code. CONTAIN 2.0 is a highly flexible and modular code that can run problems that are either quite simple or highly complex. An important aspect of CONTAIN is that the interactions among thermal-hydraulic phenomena, aerosol behavior, and fission product behavior are taken into account. The code includes atmospheric models for steam/air thermodynamics, intercell flows, condensation/evaporation on structures and aerosols, aerosol behavior, and gas combustion. It also includes models for reactor cavity phenomena such as core-concrete interactions and coolant pool boiling. Heat conduction in structures, fission product decay and transport, radioactive decay heating, and the thermal-hydraulic and fission product decontamination effects of engineered safety features are also modeled. To the extent possible, the best available models for severe accident phenomena have been incorporated into CONTAIN, but it is intrinsic to the nature of accident analysis that significant uncertainty exists regarding numerous phenomena. In those cases, sensitivity studies can be performed with CONTAIN by means of user-specified input parameters. Thus, the code can be viewed as a tool designed to assist the knowledge reactor safety analyst in evaluating the consequences of specific modeling assumptions

Code manual for CONTAIN 2.0: A computer code for nuclear reactor containment analysis

Energy Technology Data Exchange (ETDEWEB)

Murata, K.K.; Williams, D.C.; Griffith, R.O.; Gido, R.G.; Tadios, E.L.; Davis, F.J.; Martinez, G.M.; Washington, K.E. [Sandia National Labs., Albuquerque, NM (United States); Tills, J. [J. Tills and Associates, Inc., Sandia Park, NM (United States)

1997-12-01

The CONTAIN 2.0 computer code is an integrated analysis tool used for predicting the physical conditions, chemical compositions, and distributions of radiological materials inside a containment building following the release of material from the primary system in a light-water reactor accident. It can also predict the source term to the environment. CONTAIN 2.0 is intended to replace the earlier CONTAIN 1.12, which was released in 1991. The purpose of this Code Manual is to provide full documentation of the features and models in CONTAIN 2.0. Besides complete descriptions of the models, this Code Manual provides a complete description of the input and output from the code. CONTAIN 2.0 is a highly flexible and modular code that can run problems that are either quite simple or highly complex. An important aspect of CONTAIN is that the interactions among thermal-hydraulic phenomena, aerosol behavior, and fission product behavior are taken into account. The code includes atmospheric models for steam/air thermodynamics, intercell flows, condensation/evaporation on structures and aerosols, aerosol behavior, and gas combustion. It also includes models for reactor cavity phenomena such as core-concrete interactions and coolant pool boiling. Heat conduction in structures, fission product decay and transport, radioactive decay heating, and the thermal-hydraulic and fission product decontamination effects of engineered safety features are also modeled. To the extent possible, the best available models for severe accident phenomena have been incorporated into CONTAIN, but it is intrinsic to the nature of accident analysis that significant uncertainty exists regarding numerous phenomena. In those cases, sensitivity studies can be performed with CONTAIN by means of user-specified input parameters. Thus, the code can be viewed as a tool designed to assist the knowledge reactor safety analyst in evaluating the consequences of specific modeling assumptions.

Heat-pipe effect on the transport of gaseous radionuclides released from a nuclear waste container

International Nuclear Information System (INIS)

Zhou, W.; Chambre, P.L.; Pigford, T.H.; Lee, W.W.L.

1990-11-01

When an unsaturated porous medium is subjected to a temperature gradient and the temperature is sufficiently high, vadose water is heated and vaporizes. Vapor flows under its pressure gradient towards colder regions where it condenses. Vaporization and condensation produce a liquid saturation gradient, creating a capillary pressure gradient inside the porous medium. Condensate flows towards the hot end under the influence of a capillary pressure gradient. This is a heat pipe in an unsaturated porous medium. We study analytically the transport of gaseous species released from a spent-fuel waste package, as affected by a time-dependent heat pipe in an unsaturated rock. For parameter values typical of a potential repository in partially saturated fractured tuff at Yucca Mountain, we found that a heat pipe develops shortly after waste is buried, and the heat-pipe's spatial extent is time-dependent. Water vapor movements produced by the heat pipe can significantly affect the migration of gaseous radionuclides. 12 refs., 6 figs., 1 tab

Experimental Investigation of A Heat Pipe-Assisted Latent Heat Thermal Energy Storage System

Science.gov (United States)

Tiari, Saeed; Mahdavi, Mahboobe; Qiu, Songgang

2016-11-01

In the present work, different operation modes of a latent heat thermal energy storage system assisted by a heat pipe network were studied experimentally. Rubitherm RT55 enclosed by a vertical cylindrical container was used as the Phase Change Material (PCM). The embedded heat pipe network consisting of a primary heat pipe and an array of four secondary heat pipes were employed to transfer heat to the PCM. The primary heat pipe transports heat from the heat source to the heat sink. The secondary heat pipes transfer the extra heat from the heat source to PCM during charging process or retrieve thermal energy from PCM during discharging process. The effects of heat transfer fluid (HTF) flow rate and temperature on the thermal performance of the system were investigated for both charging and discharging processes. It was found that the HTF flow rate has a significant effect on the total charging time of the system. Increasing the HTF flow rate results in a remarkable increase in the system input thermal power. The results also showed that the discharging process is hardly affected by the HTF flow rate but HTF temperature plays an important role in both charging and discharging processes. The authors would like to acknowledge the financial supports by Temple University for the project.

GOTHIC code evaluation of alternative passive containment cooling features

International Nuclear Information System (INIS)

Gavrilas, M.; Todreas, E.N.; Driscoll, M.J.

1996-01-01

Reliance on passive cooling has become an important objective in containment design. Several reactor concepts have been set forth, which are equipped with entirely passively cooled containments. However, the problems that have to be overcome in rejecting the entire heat generated by a severe accident in a high-rating reactor (i.e. one with a rating greater than 1200 MW e ) have been found to be substantial and without obvious solutions. The GOTHIC code was verified and modified for containment cooling applications; optimal mesh sizes, computational time steps and applicable heat transfer correlations were examined. The effect of the break location on circulation patterns that develop inside the containment was also evaluated. The GOTHIC code was then employed to assess the effectiveness of several original heat rejection features that make it possible to cool high-rating containments. Two containment concepts were evaluated: one for a 1200 MW e new pressure tube light-water reactor, and one for a 1300 MW e pressurized-water reactor. The effectiveness of various containment configurations that include specific pressure-limiting features has been predicted. The best-performance configurations-worst-case-accident scenarios that were examined yielded peak pressures of less than 0.30 MPa for the 1200 MW e pressure tube light-water reactor, and less than 0.45 MPa for the 1300 MW e pressurized-water reactor. (orig.)

Exergetic analysis and thermoeconomic study for a container-housed engine

Energy Technology Data Exchange (ETDEWEB)

Sala, J.M.; Eguia, J.; Flores, I. [Escuela Superior de Ingenieros Industriales de Bilbao, Universidad del Pais Vasco, Alameda de Urquijo, s/n 48013 Bilbao, Bizkaia (Spain); Lopez-Gonzalez, L.M.; Ruiz de Adana, M. [Escuela Tecnica Superior de Ingenieria Industrial, Universidad de La Rioja, C/Luis de Ulloa, 20. E-26004 Logrono, La Rioja (Spain); Miguez, J.L. [Universidad de Vigo, Escuela Tecnica Superior de Ingenieros Industriales, C/Lagoas-Marcosende, s/n 36200 Vigo, Pontevedra (Spain)

2006-11-15

This paper presents the engineering design and theoretical exergetic analyses for a container-housed reciprocating engine. The exergy analysis conducted was based on the first and second laws of thermodynamics for power generation systems. Using thermographic inspection, the heat dissipated by each one of the 28 elements under consideration in the engine container was assessed, together with the mass flow rate of air supplied to the cab and the air temperature at the inlet and outlet. This information is essential for the proper design of the ventilation system needed to disseminate the heat generated inside the container-housed unit. The energy balance and corresponding exergy balance were drawn up using the values thus obtained and the information available on the engine. The engine efficiency rates were evaluated on both an energy and exergy basis, taking into account that heat from the cooling circuit and exhaust fumes are used for CHP. Finally, thermoeconomics was applied to calculate the costs of the electricity and useful heat energy produced by the engine. The result of this study should be to optimize the design of container-housed CHP systems, showing where exergy losses occur and identifying areas of improvement. (author)

Exergetic analysis and thermoeconomic study for a container-housed engine

International Nuclear Information System (INIS)

Sala, J.M.; Lopez-Gonzalez, L.M.; Ruiz de Adana, M.; Miguez, J.L.; Eguia, J.; Flores, I.

2006-01-01

This paper presents the engineering design and theoretical exergetic analyses for a container-housed reciprocating engine. The exergy analysis conducted was based on the first and second laws of thermodynamics for power generation systems. Using thermographic inspection, the heat dissipated by each one of the 28 elements under consideration in the engine container was assessed, together with the mass flow rate of air supplied to the cab and the air temperature at the inlet and outlet. This information is essential for the proper design of the ventilation system needed to disseminate the heat generated inside the container-housed unit. The energy balance and corresponding exergy balance were drawn up using the values thus obtained and the information available on the engine. The engine efficiency rates were evaluated on both an energy and exergy basis, taking into account that heat from the cooling circuit and exhaust fumes are used for CHP. Finally, thermoeconomics was applied to calculate the costs of the electricity and useful heat energy produced by the engine. The result of this study should be to optimize the design of container-housed CHP systems, showing where exergy losses occur and identifying areas of improvement

Low-temperature nuclear heat applications: Nuclear power plants for district heating

International Nuclear Information System (INIS)

1987-08-01

The IAEA reflected the needs of its Member States for the exchange of information in the field of nuclear heat application already in the late 1970s. In the early 1980s, some Member States showed their interest in the use of heat from electricity producing nuclear power plants and in the development of nuclear heating plants. Accordingly, a technical committee meeting with a workshop was organized in 1983 to review the status of nuclear heat application which confirmed both the progress made in this field and the renewed interest of Member States in an active exchange of information about this subject. In 1985 an Advisory Group summarized the Potential of Low-Temperature Nuclear Heat Application; the relevant Technical Document reviewing the situation in the IAEA's Member States was issued in 1986 (IAEA-TECDOC-397). Programme plans were made for 1986-88 and the IAEA was asked to promote the exchange of information, with specific emphasis on the design criteria, operating experience, safety requirements and specifications for heat-only reactors, co-generation plants and power plants adapted for heat application. Because of a growing interest of the IAEA's Member States about nuclear heat employment in the district heating domaine, an Advisory Group meeting was organized by the IAEA on ''Low-Temperature Nuclear Heat Application: Nuclear Power Plants for District Heating'' in Prague, Czechoslovakia in June 1986. The information gained up to 1986 and discussed during this meeting is embodied in the present Technical Document. 22 figs, 11 tabs

Effect of resins of heat exchanger fouling by asphaltene-containing oils

Energy Technology Data Exchange (ETDEWEB)

Al-Atar, E.; Watkinson, A.P. [British Columbia Univ., Dept. of Chemical and Bio-Resource Engineering, Vancouver, BC (Canada)

1999-07-01

The effects of resins on the thermal fouling of asphaltene containing oils in heat exchangers was investigated as well as the nature of the deposits. Building on previous research, a sample of de-asphalted vacuum bottoms (DAO), serving as a source of natural resins, heavy oil (HO) and fuel oil was used to investigate the effects of resin concentration on the rate of thermal fouling. The conditions of the study included: fluid circulation through the UBC annular fouling test section for up to 30 hour periods, monitoring of thermal fouling by measurement, and nitrogen atmospheres at a bulk temperature of 85 degrees C, a bulk velocity of 0.85 am/s, and a pressure of 410 kPa. Physical and chemical characterization of the deposits was affected, and filtration at the bulk temperature before and after a run was used to determine the occurrence of fine solids in the fluid. The rate of fouling generally decreased tending generally towards asymptotic behavior in the limit, and after one day Rf values up to 0.3 m2K/kW occurred with severe fouling. An increase in the fouling rate occurred with increased DAO concentration in the mixture, at a fixed heavy oil concentration of 5 weight percent, and the relation between Asomaning's colloidal instability index and the trends in fouling rate was not observed, although there were some indications of reduced fouling as there was an increase in the ratio of resins to asphaltenes, however, blends of the DAO-HO-FO helped to control the concentration of asphaltenes and resins that are possible. (Abstract only).

Children's Literature: An Issues Approach.

Science.gov (United States)

Rudman, Masha Kabakow

Meant as a reference and guide to critical evaluation of the way issues are treated in children's books, each of the nine chapters in this book can be used independently. Each chapter contains a discussion of an issue, a section that relates how particular books handle that issue, at least two suggestions for activities that teachers can use to…

Using geothermal energy to heat a portion of a formation for an in situ heat treatment process

Science.gov (United States)

Pieterson, Roelof; Boyles, Joseph Michael; Diebold, Peter Ulrich

2010-06-08

Methods of using geothermal energy to treat subsurface formations are described herein. Methods for using geothermal energy to treat a subsurface treatment area containing or proximate to hydrocarbons may include producing geothermally heated fluid from at least one subsurface region. Heat from at least a portion of the geothermally heated fluid may be transferred to the subsurface treatment area to heat the subsurface treatment area. At least some hydrocarbon fluids may be produced from the formation.

Heat exchanger

Science.gov (United States)

Wolowodiuk, Walter

1976-01-06

A heat exchanger of the straight tube type in which different rates of thermal expansion between the straight tubes and the supply pipes furnishing fluid to those tubes do not result in tube failures. The supply pipes each contain a section which is of helical configuration.

Testing of PCM Heat Storage Modules with Solar Collectors as Heat Source

DEFF Research Database (Denmark)

Englmair, Gerald; Dannemand, Mark; Johansen, Jakob Berg

2016-01-01

A latent heat storage based on the phase change material Sodium Acetate Trihydrate (SAT) has been tested as part of a demonstration system. The full heat storage consisted of 4 individual modules each containing about 200 kg of sodium acetate trihydrate with different additives. The aim...... was to actively utilize the ability of the material to supercool to obtain long storage periods. The modules were charged with solar heat supplied by 22.4 m2 evacuated tubular collectors. The investigation showed that it was possible to fully charge one module within a period of 270 minutes with clear skies...

Special issue on electron cyclotron wave physics, technology, and applications - Part 2

International Nuclear Information System (INIS)

Uckan, Nermin A.

2008-01-01

This issue of Fusion Science and Technology (FS and T) contains a compendium of full-length, peer-reviewed papers on electron cyclotron (EC) wave physics, technology, and applications on magnetically confined plasmas. The interest in this special issue started with a simple question from a single individual who asked if he could submit for publication in FS and T his paper ''ITER ECH Front Steering Upper Launcher,'' parts of which he was planning to present at the 14th Joint Workshop on Electron Cyclotron Emission and Electron Cyclotron Resonance Heating, Santorini Island, Greece, May 2006. Such interest quickly grew, and the decision was made to offer the same opportunity to other workshop participants as well as to other interested researchers from around the world to contribute to a special FS and T issue on EC wave physics, technology, and applications. The person who started this ''wave'' of interest is no other than Dr. Mark Henderson, who was later drafted and kindly agreed to serve as the guest editor for this issue. The worldwide research program on EC wave physics, technology, and applications has shown impressive progress over the past couple of years, and much of this progress is reflected in the fifty or so papers that are included in this two-part special issue - part 1 in August 2007 and part 2 in January 2008. To complement the contributed papers, several informative reviews, which will be valuable for years to come, were also invited and are included. These review papers provide an objective summary of the current state of the art in EC emission research, theory of EC waves, EC heating and current drive experiments, gyrotron development, launcher development, and transmission systems. In preparation for ITER, this special issue is timely and should be of interest to those already working in the field and to the new generation of scientists and engineers who will be the ones to design, build, and carry out experiments on ITER. We extend our

Simulation of the heat transfer of a irradiated fuel storage container with code CFD STAR- CCM+; Simulacion de la transferencia de calor de un contenedor de almacenamiento de combustible irradiado con el codigo CFD STAR-CCM+

Energy Technology Data Exchange (ETDEWEB)

Barrera matalla, J. E.; Hernandez Gomez, J.; Riverala Gurruchaga, J.

2012-07-01

Irradiated fuel has become an object of interest in the industry by the importance of ensuring its safety during long periods of storage time. New containers, stores, methods and codes will be used to ensure a suitable cooling and residual heat removal, and secure the safety of fuel elements in dry storage. The codes CFD (Computational Fluid Dynamics) have great potential to help in design of containers and stores, improving thermal-hydraulic performance and the extraction of heat generated.

Prediction of Heat Removal Capacity of Horizontal Condensation Heat Exchanger submerged in Pool

Energy Technology Data Exchange (ETDEWEB)

Jeon, Seong-Su; Hong, Soon-Joon [FNC Tech., Yongin (Korea, Republic of); Cho, Hyoung-Kyu [Seoul National University, Seoul (Korea, Republic of); Park, Goon-Cherl [KEPCO International Nuclear Graduate School, Ulsan (Korea, Republic of)

2014-10-15

As representative passive safety systems, there are the passive containment cooling system (PCCS) of ESBWR, the emergency condenser system (ECS) of the SWR-1000, the passive auxiliary feed-water system (PAFS) of the APR+ and etc. During the nuclear power plant accidents, these passive safety systems can cool the nuclear system effectively via the heat transfer through the steam condensation, and then mitigate the accidents. For the optimum design and the safety analysis of the passive safety system, it is essential to predict the heat removal capacity of the heat exchanger well. The heat removal capacity of the horizontal condensation heat exchanger submerged in a pool is determined by a combination of a horizontal in-tube condensation heat transfer and a boiling heat transfer on the horizontal tube. Since most correlations proposed in the previous nuclear engineering field were developed for the vertical tube, there is a certain limit to apply these correlations to the horizontal tube. Therefore, this study developed the heat transfer model for the horizontal Ushaped condensation heat exchanger submerged in a pool to predict well the horizontal in-tube condensation heat transfer, the boiling heat transfer on the horizontal tube and the overall heat removal capacity of the heat exchanger using the best-estimate system analysis code, MARS.

EDITORIAL: Special issue containing contributions from the Fourth International Brain-Computer Interface Meeting Special issue containing contributions from the Fourth International Brain-Computer Interface Meeting

Science.gov (United States)

Vaughan, Theresa M.; Wolpaw, Jonathan R.

2011-04-01

This special issue of Journal of Neural Engineering is a result of the Fourth International Brain-Computer Interface Meeting, which was held at the Asilomar Conference Center in Monterey, California, USA from 31 May to 4 June, 2010. The meeting was sponsored by the National Institutes of Health, The National Science Foundation and the Department of Defense, and was organized by the Wadsworth Center of the New York State Department of Health. It attracted over 260 participants from 17 countries—including many graduate students and postdoctoral fellows—and featured 19 workshops, platform presentations from 26 research groups, 170 posters, multiple brain-computer interface (BCI) demonstrations, and a keynote address by W Zev Rymer of the Rehabilitation Institute of Chicago. The number of participants and the diversity of the topics covered greatly exceeded those of the previous meeting in 2005, and testified to the continuing rapid expansion and growing sophistication of this exciting and still relatively new research field. BCI research focuses primarily on using brain signals to replace or restore the motor functions that people have lost due to amyotrophic lateral sclerosis (ALS), a brainstem stroke, or some other devastating neuromuscular disorder. In the last few years, attention has also turned towards using BCIs to improve rehabilitation after a stroke, and beyond that to enhancing or supplementing the capabilities of even those without disabilities. These diverse interests were represented in the wide range of topics covered in the workshops. While some workshops addressed broad traditional topics, such as signal acquisition, feature extraction and translation, and software development, many addressed topics that were entirely new or focused sharply on areas that have become important only recently. These included workshops on optimizing P300-based BCIs; improving the mutual adaptations of the BCI and the user; BCIs that can control neuroprostheses

Transport containers for radioactive material

International Nuclear Information System (INIS)

Bibby, D.

1978-01-01

A transport container for transporting irradiated nuclear fuel is described comprising a steel flask with detachable cover and having external heat exchange fins. The flask contains a solid annular shield comprised of discrete bodies of Pb or Fe bonded together by a solid matrix, for attenuating gamma rays and neutron emission. This may comprise lead shot bonded together by concrete or polyethylene, or alternatively iron shot bonded by concrete. (UK)

Immobilization of radwastes in glass containers and products formed thereby

International Nuclear Information System (INIS)

Macedo, P.B.; Litovitz, T.A.; Simmons, C.J.; Simmons, J.H.; Lagakos, N.; Tran, D.C.

1982-01-01

A mixture of glass packing and radioactive or other toxic material is placed in a solid glass container, and the container is heated to drive off volatile components and collapse and seal the container

"Nanotechnology Enabled Advanced Industrial Heat Transfer Fluids"

Energy Technology Data Exchange (ETDEWEB)

Dr. Ganesh Skandan; Dr. Amit Singhal; Mr. Kenneth Eberts; Mr. Damian Sobrevilla; Prof. Jerry Shan; Stephen Tse; Toby Rossmann

2008-06-12

ABSTRACT Nanotechnology Enabled Advanced industrial Heat Transfer Fluids” Improving the efficiency of Industrial Heat Exchangers offers a great opportunity to improve overall process efficiencies in diverse industries such as pharmaceutical, materials manufacturing and food processing. The higher efficiencies can come in part from improved heat transfer during both cooling and heating of the material being processed. Additionally, there is great interest in enhancing the performance and reducing the weight of heat exchangers used in automotives in order to increase fuel efficiency. The goal of the Phase I program was to develop nanoparticle containing heat transfer fluids (e.g., antifreeze, water, silicone and hydrocarbon-based oils) that are used in transportation and in the chemical industry for heating, cooling and recovering waste heat. Much work has been done to date at investigating the potential use of nanoparticle-enhanced thermal fluids to improve heat transfer in heat exchangers. In most cases the effect in a commercial heat transfer fluid has been marginal at best. In the Phase I work, we demonstrated that the thermal conductivity, and hence heat transfer, of a fluid containing nanoparticles can be dramatically increased when subjected to an external influence. The increase in thermal conductivity was significantly larger than what is predicted by commonly used thermal models for two-phase materials. Additionally, the surface of the nanoparticles was engineered so as to have a minimal influence on the viscosity of the fluid. As a result, a nanoparticle-laden fluid was successfully developed that can lead to enhanced heat transfer in both industrial and automotive heat exchangers

Heat pipes. (Latest citations from the US Patent Bibliographic file with exemplary claims). Published Search

Energy Technology Data Exchange (ETDEWEB)

NONE

1995-01-01

The bibliography contains citations of selected patents concerning the design, manufacture, and applications of heat pipes. The use of heat pipes in heat exchange systems for heat storage, heat transfer, and heat utilization is discussed. Applications include semiconductor cooling, use in engine components, and building cooling and heating. (Contains 250 citations and includes a subject term index and title list.)

Fusion Canada issue 16

Energy Technology Data Exchange (ETDEWEB)

NONE

1992-01-01

A short bulletin from the National Fusion Program. Included in this issue is a report on Ontario Hydro`s contract for tritium supply from Germany, a CCFM update on plasma biasing, divertor operation, radiofrequency plasma current drive and the plasma heating system for the Tokamak de Varennes, and an agreement for ITER engineering design activities. 5 figs.

Fusion Canada issue 16

International Nuclear Information System (INIS)

1992-01-01

A short bulletin from the National Fusion Program. Included in this issue is a report on Ontario Hydro's contract for tritium supply from Germany, a CCFM update on plasma biasing, divertor operation, radiofrequency plasma current drive and the plasma heating system for the Tokamak de Varennes, and an agreement for ITER engineering design activities. 5 figs

Occupational Heat Stress Profiles in Selected Workplaces in India.

Science.gov (United States)

Venugopal, Vidhya; Chinnadurai, Jeremiah S; Lucas, Rebekah A I; Kjellstrom, Tord

2015-12-29

Health and productivity impacts from occupational heat stress have significant ramifications for the large workforce of India. This study profiled occupational heat stress impacts on the health and productivity of workers in select organized and unorganized Indian work sectors. During hotter and cooler seasons, Wet Bulb Globe Temperatures (WBGT) were used to quantify the risk of heat stress, according to International workplace guidelines. Questionnaires assessed workers' perceived health and productivity impacts from heat stress. A total of 442 workers from 18 Indian workplaces participated (22% and 78% from the organized and unorganized sector, respectively). Overall 82% and 42% of workers were exposed to higher than recommended WBGT during hotter and cooler periods, respectively. Workers with heavy workloads reported more heat-related health issues (chi square = 23.67, p ≤ 0.001) and reduced productivity (chi square = 15.82, p ≤ 0.001), especially the outdoor workers. Heat-rashes, dehydration, heat-syncope and urinogenital symptoms were self-reported health issues. Cited reasons for productivity losses were: extended-work hours due to fatigue/exhaustion, sickness/hospitalization and wages lost. Reducing workplace heat stress will benefit industries and workers via improving worker health and productivity. Adaptation and mitigation measures to tackle heat stress are imperative to protect the present and future workforce as climate change progresses.

Occupational Heat Stress Profiles in Selected Workplaces in India

Science.gov (United States)

Venugopal, Vidhya; Chinnadurai, Jeremiah S.; Lucas, Rebekah A. I.; Kjellstrom, Tord

2015-01-01

Health and productivity impacts from occupational heat stress have significant ramifications for the large workforce of India. This study profiled occupational heat stress impacts on the health and productivity of workers in select organized and unorganized Indian work sectors. During hotter and cooler seasons, Wet Bulb Globe Temperatures (WBGT) were used to quantify the risk of heat stress, according to International workplace guidelines. Questionnaires assessed workers’ perceived health and productivity impacts from heat stress. A total of 442 workers from 18 Indian workplaces participated (22% and 78% from the organized and unorganized sector, respectively). Overall 82% and 42% of workers were exposed to higher than recommended WBGT during hotter and cooler periods, respectively. Workers with heavy workloads reported more heat-related health issues (chi square = 23.67, p ≤ 0.001) and reduced productivity (chi square = 15.82, p ≤ 0.001), especially the outdoor workers. Heat-rashes, dehydration, heat-syncope and urinogenital symptoms were self-reported health issues. Cited reasons for productivity losses were: extended-work hours due to fatigue/exhaustion, sickness/hospitalization and wages lost. Reducing workplace heat stress will benefit industries and workers via improving worker health and productivity. Adaptation and mitigation measures to tackle heat stress are imperative to protect the present and future workforce as climate change progresses. PMID:26729144

Catalytic heat exchangers - a long-term evaluation

Energy Technology Data Exchange (ETDEWEB)

Silversand, Fredrik A. [CATATOR AB, Lund (Sweden)

2003-10-01

A long-term evaluation concerning catalytic heat exchangers (CHEs) has been performed. The idea concerning CHEs was originally described in a number of reports issued by Catator almost a decade ago. The general idea with CHEs is to combust a fuel with a catalyst inside a heat exchanger to enable an effective heat transfer. The first design approaches demonstrated the function and the possibilities with CHEs but were defective concerning the heat exchanger design. Consequently, a heat exchanger company (SWEP International AB), which was specialised on brazed plate-type heat exchangers, joined the continued development project. Indeed, the new design approach containing Catator's wire-mesh catalysts and SWEP's plate-type heat exchangers enabled us to improve the concept considerably. The new design complied with a number of relevant technical demands, e.g.: Simplicity; Compactness and integration (few parts); High thermal efficiency; Low pressure drop; Excellent emissions; High turn-down ratio; Reasonable production cost. Spurred by the technical progresses, the importance of a long-term test under realistic conditions was clear. A long-term evaluation was initialised at Sydkraft Gas premises in Aastorp. The CHE was installed on a specially designed rig to enable accelerated testing with respect to the number of transients. The rig was operated continuously for 5000 hours and emission mapping was carried out at certain time intervals. Following some problems during the initial phase of the long-term evaluation, which unfortunately also delayed the project, the results indicated very stable conditions of operation. The emissions have been rather constant during the course of the test and we cannot see any tendencies to decreased performances. Indeed, the test verifies the function, operability and reliability of the CHE-concept. Apart from domestic boilers we foresee a number of interesting and relevant applications in heating and process technology. Since

Analysis code for pressure in reactor containment vessel of ATR. CONPOL

International Nuclear Information System (INIS)

1997-08-01

For the evaluation of the pressure and temperature in containment vessels in the events which are classified in the abnormal change of pressure, atmosphere and others in reactor containment vessels in accident among the safety evaluation events of the ATR, the analysis code for the pressure in reactor containment vessels CONPOL is used. In this report, the functions of the analysis code and the analysis model are shown. By using this analysis code, the rise of the pressure and temperature in a containment vessel is evaluated when loss of coolant accident occurs, and high temperature, high pressure coolant flows into it. This code possesses the functions of computing blow-down quantity and heat dissipation from reactor cooling facility, steam condensing heat transfer to containment vessel walls, and the cooling effect by containment vessel spray system. As for the analysis techniques, the models of reactor cooling system, containment vessel and steam discharge pool, and the computation models for the pressure and temperature in containment vessels, wall surface temperature, condensing heat transfer, spray condensation and blow-down are explained. The experimental analysis of the evaluation of the pressure and temperature in containment vessels at the time of loss of coolant accident is reported. (K.I.)

Heat exchanger

International Nuclear Information System (INIS)

Wolowodiuk, W.

1976-01-01

A heat exchanger of the straight tube type is described in which different rates of thermal expansion between the straight tubes and the supply pipes furnishing fluid to those tubes do not result in tube failures. The supply pipes each contain a section which is of helical configuration

Research status and issues of tungsten plasma facing materials for ITER and beyond

International Nuclear Information System (INIS)

Ueda, Y.; Coenen, J.W.; De Temmerman, G.; Doerner, R.P.; Linke, J.; Philipps, V.; Tsitrone, E.

2014-01-01

This review summarizes surface morphology changes of tungsten caused by heat and particle loadings from edge plasmas, and their effects on enhanced erosion and material lifetime in ITER and beyond. Pulsed heat loadings by transients (disruption and ELM) are the largest concerns due to surface melting, cracking, and dust formation. Hydrogen induced blistering is unlikely to be an issue of ITER. Helium bombardment would cause surface morphology changes such as W fuzz, He holes, and nanometric bubble layers, which could lead to enhanced erosion (e.g. unipolar arcing of W fuzz). Particle loadings could enhance pulsed heat effects (cracking and erosion) due to surface layer embrittlement by nanometric bubbles and solute atoms. But pulsed heat loadings alleviate surfaces morphology changes in some cases (He holes by ELM-like heat pulses). Effects of extremely high fluence (∼10 30 m −2 ), mixed materials, and neutron irradiation are important issues to be pursued for ITER and beyond. In addition, surface refurbishment to prolong material lifetime is also an important issue

Enhancing heat capacity of colloidal suspension using nanoscale encapsulated phase-change materials for heat transfer.

Science.gov (United States)

Hong, Yan; Ding, Shujiang; Wu, Wei; Hu, Jianjun; Voevodin, Andrey A; Gschwender, Lois; Snyder, Ed; Chow, Louis; Su, Ming