Fast Generation of Container Vessel Stowage Plans

DEFF Research Database (Denmark)

Pacino, Dario

that the vessel is stable and seaworthy, and at the same time arrange the cargo such that the time at port is minimized. Moreover, stowage coordinators only have a limited amount of time to produce the plan. This thesis addresses the question of whether it is possible to automatically generate stowage plans...... test instances provided by a major liner shipping company. Improvements to the modeling of vessel stability and an analysis of its accuracy together with an analysis of the computational complexity of the container stowage problem are also included in the thesis, resulting in an overall in...

Mark III Containment vessel/annulus concrete design

International Nuclear Information System (INIS)

Chang, P.S.; Moussa, M.M.

1981-01-01

Recently, engineers have been considering the significant dynamic impact of safety/relief valve (S/RV) discharge loads on the containment structures, safety equipment, and piping systems in BWR type reactors. For a plant in the construction stage, extensive modifications will be made to qualify these new loads. The lower portion of the containment vessel serves as a suppression pool pressure boundary and is designed to sustain the effects of postulated loss of coolant accidents, seismic occurrences, S/RV discharge loads, and other effects. Extremely high spectral peak accelerations of the free-standing steel containment vessel can be obtained during the air dearing process of the S/RV discharge. Parametric studies indicated that a substantial reduction in response can be obtained by increasing the stiffness of the steel containment vessel in the lover area. A concrete backing configuration in the suppression pool area of Mark III Containment is proposed in this paper. A composite action is assumed between the steel containment vessel shell and the concrete section. The system is physically separated from the shield building. This approach warrants an early erection of the shield building and a late installation of piping systems in the containment vessel suppression pool area. Finite element analyses are performed by using ASHSD2 and EASE2 computer codes. The results of the analyses have shown the proposed stress criteria are satisfied. The approach pressented is justified to be a workable system for a new plant design. (orig./HP)

Round Robin Posttest analysis of a 1/10-scale Steel Containment Vessel Model Test

International Nuclear Information System (INIS)

Komine, Kuniaki; Konno, Mutsuo

1999-01-01

NUPEC and U.S. Nuclear Regulatory Commission (USNRC) have been jointly sponsoring 'Structural Behavior Test' at Sandia National Laboratory (SNL) in Cooperative Containment Research Program'. As one of the test, a test of a mixed scaled SCV model with 1/10 in the geometry and 1/4 in the shell thickness. Round Robin analyses of a 1/10-scale Steel Containment Vessel (SCV) Model Test were carried out to obtain an adequate analytical method among seven organizations belonged to five countries in the world. As one of sponsor, Nuclear Power Engineering Corporation (NUPEC) filled the important role of a posttest analysis of SCV model. This paper describes NUPEC's analytical results in the round robin posttest analysis. (author)

Safety analysis of nuclear containment vessels subjected to strong earthquakes and subsequent tsunamis

Energy Technology Data Exchange (ETDEWEB)

Lin, Feng; Li, Hong Zhi [Dept. Structural Engineering, Tongji University, Shanghai (China)

2017-08-15

Nuclear power plants under expansion and under construction in China are mostly located in coastal areas, which means they are at risk of suffering strong earthquakes and subsequent tsunamis. This paper presents a safety analysis for a new reinforced concrete containment vessel in such events. A finite element method-based model was built, verified, and first used to understand the seismic performance of the containment vessel under earthquakes with increased intensities. Then, the model was used to assess the safety performance of the containment vessel subject to an earthquake with peak ground acceleration (PGA) of 0.56g and subsequent tsunamis with increased inundation depths, similar to the 2011 Great East earthquake and tsunami in Japan. Results indicated that the containment vessel reached Limit State I (concrete cracking) and Limit State II (concrete crushing) when the PGAs were in a range of 0.8–1.1g and 1.2–1.7g, respectively. The containment vessel reached Limit State I with a tsunami inundation depth of 10 m after suffering an earthquake with a PGA of 0.56g. A site-specific hazard assessment was conducted to consider the likelihood of tsunami sources.

Containment vessel construction for nuclear power reactors

International Nuclear Information System (INIS)

Sulzer, H.D.; Coletti, J.L.

1975-01-01

A nuclear containment vessel houses an inner reactor housing structure whose outer wall is closely spaced from the inner wall of the containment vessel. The inner reactor housing structure is divided by an intermediate floor providing an upper chamber for housing the reactor and associated steam generators and a lower chamber directly therebeneath containing a pressure suppression pool. Communication between the upper chamber and the pressure suppression pool is established by conduits extending through the intermediate floor which terminate beneath the level of the pressure suppression pool and by inlet openings in the reactor housing wall beneath the level of the pressure suppression pool which communicate with the annulus formed between the outer wall of the reactor housing structure and the inner wall of the containment vessel. (Official Gazette)

Round Robin Posttest analysis of a 1/10-scale Steel Containment Vessel Model Test

Energy Technology Data Exchange (ETDEWEB)

Komine, Kuniaki [Nuclear Power Engineering Corp., Tokyo (Japan); Konno, Mutsuo

1999-07-01

NUPEC and U.S. Nuclear Regulatory Commission (USNRC) have been jointly sponsoring 'Structural Behavior Test' at Sandia National Laboratory (SNL) in Cooperative Containment Research Program'. As one of the test, a test of a mixed scaled SCV model with 1/10 in the geometry and 1/4 in the shell thickness. Round Robin analyses of a 1/10-scale Steel Containment Vessel (SCV) Model Test were carried out to obtain an adequate analytical method among seven organizations belonged to five countries in the world. As one of sponsor, Nuclear Power Engineering Corporation (NUPEC) filled the important role of a posttest analysis of SCV model. This paper describes NUPEC's analytical results in the round robin posttest analysis. (author)

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

International Nuclear Information System (INIS)

Porter, V.L.

1994-01-01

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

Safety analysis of nuclear containment vessels subjected to strong earthquakes and subsequent tsunamis

Directory of Open Access Journals (Sweden)

Feng Lin

2017-08-01

Full Text Available Nuclear power plants under expansion and under construction in China are mostly located in coastal areas, which means they are at risk of suffering strong earthquakes and subsequent tsunamis. This paper presents a safety analysis for a new reinforced concrete containment vessel in such events. A finite element method-based model was built, verified, and first used to understand the seismic performance of the containment vessel under earthquakes with increased intensities. Then, the model was used to assess the safety performance of the containment vessel subject to an earthquake with peak ground acceleration (PGA of 0.56g and subsequent tsunamis with increased inundation depths, similar to the 2011 Great East earthquake and tsunami in Japan. Results indicated that the containment vessel reached Limit State I (concrete cracking and Limit State II (concrete crushing when the PGAs were in a range of 0.8–1.1g and 1.2–1.7g, respectively. The containment vessel reached Limit State I with a tsunami inundation depth of 10 m after suffering an earthquake with a PGA of 0.56g. A site-specific hazard assessment was conducted to consider the likelihood of tsunami sources.

Stowing the Right Containers on Container Vessels

DEFF Research Database (Denmark)

Jensen, Rune Møller

2014-01-01

’s largest container vessels using standard mathematical programming techniques and off-the-shelf solvers. The presentation will provide basic insight into the domain, with pointers to further information that enable you to join in this promising new path of operations research and business....

Preliminary calculation with code CONTEMPT-LT for spray cooling tests with JAERI model containment vessel

International Nuclear Information System (INIS)

Tanaka, Mitsugu

1978-01-01

LWR plants have a containment spray system to reduce the escape of radioactive material to the environment in a loss-of-coolant accident (LOCA) by washing out fission products, especially radioiodine, and condensing the steam to lower the pressure. For carrying out the containment spray tests, pressure and temperature behaviour of the JAERI Model Containment Vessel in spray cooling has been calculated with computer program CONTEMPT-LT. The following could be studied quantitatively: (1) pressure and temperature raise rates for steam addition rate and (2) pressure fall rate for spray flow rate and spray heat transfer efficiency. (auth.)

The measured contribution of whipping and springing on the fatigue and extreme loading of container vessels

Science.gov (United States)

Storhaug, Gaute

2014-12-01

Whipping/springing research started in the 50'ies. In the 60'ies inland water vessels design rules became stricter due to whipping/springing. The research during the 70-90'ies may be regarded as academic. In 2000 a large ore carrier was strengthened due to severe cracking from North Atlantic operation, and whipping/springing contributed to half of the fatigue damage. Measurement campaigns on blunt and slender vessels were initiated. A few blunt ships were designed to account for whipping/springing. Based on the measurements, the focus shifted from fatigue to extreme loading. In 2005 model tests of a 4,400 TEU container vessel included extreme whipping scenarios. In 2007 the 4400 TEU vessel MSC Napoli broke in two under similar conditions. In 2009 model tests of an 8,600 TEU container vessel container vessel included extreme whipping scenarios. In 2013 the 8,100 TEU vessel MOL COMFORT broke in two under similar conditions. Several classification societies have published voluntary guidelines, which have been used to include whipping/springing in the design of several container vessels. This paper covers results from model tests and full scale measurements used as background for the DNV Legacy guideline. Uncertainties are discussed and recommendations are given in order to obtain useful data. Whipping/springing is no longer academic.

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

International Nuclear Information System (INIS)

Porter, V.L.

1993-01-01

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

Concrete containment vessels (CCV) for nuclear power plants, (1)

International Nuclear Information System (INIS)

Ibe, Yukimi; Kitajima, Masatake

1977-01-01

Containment vessels (CV) and the construction of concrete containment vessels (CCV) for nuclear power plants are described generally, and their use and techniques in foreign countries are illustrated, in connection with the introduction of CCV to Japanese nuclear power plants. The introduction deals with the construction plan of Japanese nuclear power plants, and with the difficulties in the steel CV for large scale construction. The investigations, tests and researches are not yet sufficient. The prompt establishment of safety supported by technical criteria, analytical methods and experiments is desired. The second part deals with the consideration for aseismatic design, construction, function and characteristics of CCV. The classification and currently employed CCV, which is mainly reinforced concrete containment vessels (RCCV), are described, and the typical CCV employed for BWR is illustrated. Further, the typical arrangement of reinforcing steels at the cylindrical portion and the dome portion of RCCV is illustrated. The third part deals with the present state of CCV abroad. A prestressed concrete containment vessel (PCCV) of Turkey Point power plant is illustrated as a typical example of CCV. The tests reported in the international meeting for the design, construction and operation of concrete pressure vessels and concrete containment vessels at York University in England in 1975 are reviewed. Typical examples of the design conditions, the size and form, and the construction procedure for PCCV and RCCV abroad are reviewed. (Iwakiri, K.)

Evaluation of buckling on containment metallic vessels

International Nuclear Information System (INIS)

Silveira, Renato Campos da; Mattar Neto, Miguel

2000-01-01

The buckling analysis represents one of the most important aspects of the structural projects of nuclear power plants containment metallic vessels and in this work the Case N-284-1 ASME Code is used for evaluation of those structures submitted to this failure mode. From the stress analysis, performed by using finite element method on discrete structures with shell elements, the procedure of the Code Case are applied to the evaluation of the containment metallic vessel of the Angra 2 nuclear power plant submitted to the own weight, seismic loads and uplift in case of accident. A study of pressure vessel reinforced by rings submit ed to the external pressure. Conclusions and commentaries are established based on the obtained results

Proof testing of an explosion containment vessel

Energy Technology Data Exchange (ETDEWEB)

Esparza, E.D. [Esparza (Edward D.), San Antonio, TX (United States); Stacy, H.; Wackerle, J. [Los Alamos National Lab., NM (United States)

1996-10-01

A steel containment vessel was fabricated and proof tested for use by the Los Alamos National Laboratory at their M-9 facility. The HY-100 steel vessel was designed to provide total containment for high explosives tests up to 22 lb (10 kg) of TNT equivalent. The vessel was fabricated from an 11.5-ft diameter cylindrical shell, 1.5 in thick, and 2:1 elliptical ends, 2 in thick. Prior to delivery and acceptance, three types of tests were required for proof testing the vessel: a hydrostatic pressure test, air leak tests, and two full design charge explosion tests. The hydrostatic pressure test provided an initial static check on the capacity of the vessel and functioning of the strain instrumentation. The pneumatic air leak tests were performed before, in between, and after the explosion tests. After three smaller preliminary charge tests, the full design charge weight explosion tests demonstrated that no yielding occurred in the vessel at its rated capacity. The blast pressures generated by the explosions and the dynamic response of the vessel were measured and recorded with 33 strain channels, 4 blast pressure channels, 2 gas pressure channels, and 3 displacement channels. This paper presents an overview of the test program, a short summary of the methodology used to predict the design blast loads, a brief description of the transducer locations and measurement systems, some of the hydrostatic test strain and stress results, examples of the explosion pressure and dynamic strain data, and some comparisons of the measured data with the design loads and stresses on the vessel.

Method of burying vessel containing radioactive waste

International Nuclear Information System (INIS)

Koga, Yoshihito.

1989-01-01

A float having an inert gas sealed therein is attached to a tightly closed vessel containing radioactive wastes. The vessel is inserted and kept in a small hole for burying the tightly closed vessel in an excavated shaft in rocks such as of granite or rock salts, while filling bentonite as shielding material therearound. In this case, the float is so adjusted that the apparent specific gravity is made equal or nearer between the tightly closed vessel and the bentonite, so that the rightly closed vessel does not sink and cause direct contact with the rocks even if bentonite flows due to earthquakes, etc. This can prevent radioactivity contamination through water in the rocks. (S.K.)

Seismic proving test of PWR reactor containment vessel

International Nuclear Information System (INIS)

Akiyama, H.; Yoshikawa, T.; Tokumaru, Y.

1987-01-01

The seismic reliability proving tests of nuclear power plant facilities are carried out by Nuclear Power Engineering Test Center (NUPEC), using the large-scale, high-performance vibration of Tadotsu Engineering Laboratory, and sponsored by the Ministry of International Trade and Industry (MITI). In 1982, the seismic reliability proving test of PWR containment vessel started using the test component of reduced scale 1/3.7 and the test component proved to have structural soundness against earthquakes. Subsequently, the detailed analysis and evaluation of these test results were carried out, and the analysis methods for evaluating strength against earthquakes were established. Whereupon, the seismic analysis and evaluation on the actual containment vessel were performed by these analysis methods, and the safety and reliability of the PWR reactor containment vessel were confirmed

Sealing method and sealing device for radioactive waste containing vessel

International Nuclear Information System (INIS)

Ishiwatari, Koji; Otsuki, Akira

1998-01-01

A radioactive waste-containing body is hoisted down into a strong-material vessel opened upwardly, and a strong-material lid is hoisted down to the opening of the strong-material-vessel and welded. The strong material vessel is hoisted up and loaded on a corrosion resistant-material bottom plate placed horizontally. A corrosion resistant-material vessel having one opening end and having a corrosion resistant-material flange on the other end and previously agreed with the strong material-vessel main body is hoisted up by a hoisting device having an inserting device so that the opening of the corrosion resistant vessel is directed downwardly. The corrosion resistant vessel is press-fitted to the outside of the strong material-vessel by the inserting device while being heated by a preheater to shrink. Subsequently, the lower end of the corrosion resistant-material vessel and the corrosion resistant-material bottom plate are welded to constitute a corrosion resistant-material vessel. Then, the radioactive waste containing body can be sealed in a sealing vessel comprising the strong-material vessel and the corrosion resistant-material vessel. (N.H.)

Biaxial Loading Tests for steel containment vessel

Energy Technology Data Exchange (ETDEWEB)

Miyagawa, T. [Nuclear Power Engineering Corp., Tokyo (Japan); Wright, D.J.; Arai, S.

1999-07-01

The Nuclear Power Engineering Corporation (NUPEC) has conducted a 1/10 scale of the steel containment vessel (SCV) test for the understanding of ultimate structural behavior beyond the design pressure condition. Biaxial Loading Tests were supporting tests for the 1/10 scale SCV model to evaluate the method of estimating failure conditions of thin steel plates under biaxial loading conditions. The tentative material models of SGV480 and SPV490 were obtained. And the behavior of SGV480 and SPV490 thin steel plates under biaxial loading conditions could be well simulated by FE-Analyses with the tentative material models and Mises constitutive law. This paper describes the results and the evaluations of these tests. (author)

Biaxial Loading Tests for steel containment vessel

International Nuclear Information System (INIS)

Miyagawa, T.; Wright, D.J.; Arai, S.

1999-01-01

The Nuclear Power Engineering Corporation (NUPEC) has conducted a 1/10 scale of the steel containment vessel (SCV) test for the understanding of ultimate structural behavior beyond the design pressure condition. Biaxial Loading Tests were supporting tests for the 1/10 scale SCV model to evaluate the method of estimating failure conditions of thin steel plates under biaxial loading conditions. The tentative material models of SGV480 and SPV490 were obtained. And the behavior of SGV480 and SPV490 thin steel plates under biaxial loading conditions could be well simulated by FE-Analyses with the tentative material models and Mises constitutive law. This paper describes the results and the evaluations of these tests. (author)

Containment vessel design and practice

International Nuclear Information System (INIS)

Bangash, Y.

1983-01-01

The state of the art of analysis and design of the concrete containment vessels required for BWR and PWR is reviewed. A step-by-step critical appraisal of the existing work is given. Elastic, inelastic and cracking conditions under extreme loads are fully discussed. Problems associated with these structures are highlighted. A three-dimensional finite element analysis is included to cater for service, overload and dynamic cracking of such structures. Missile impact and seismic effects are included in this work. The second analysis is known as the limit state analysis, which is given to design such vessels for any kind of load. (U.K.)

Application of high strength steel to nuclear reactor containment vessel

International Nuclear Information System (INIS)

Susukida, H.; Sato, M.; Takano, G.; Uebayashi, T.; Yoshida, K.

1976-01-01

Nuclear reactor containment vessels are becoming larger in size with the increase in the power generating capacity of nuclear power plants. For example, a containment vessel for a PWR power plant with an output of 1,000 MWe becomes an extremely large one if it is made of the conventional JIS SGV 49 (ASTM A 516 Gr. 70) steel plates less than 38 mm in thickness. In order to design the steel containment vessel within the conventional dimensional range, therefore, it is necessary to use a high strength steel having a higher tensile strength than SGV 49 steel, good weldability and a higher fracture toughness and moreover, possessing satisfactory properties without undergoing post-weld heat treatment. The authors conducted a series of verification tests on high strength steel developed by modifying the ASTM A 543 Grade B Class 1 steel with a view to adopting it as a material for the nuclear reactor containment vessels. As the result of evaluation of the test results from various angles, we confirmed that the high strength steel is quite suitable for the manufacture of nuclear reactor containment vessels. (auth.)

Float level switch for a nuclear power plant containment vessel

International Nuclear Information System (INIS)

Powell, J.G.

1993-01-01

This invention is a float level switch used to sense rise or drop in water level in a containment vessel of a nuclear power plant during a loss of coolant accident. The essential components of the device are a guide tube, a reed switch inside the guide tube, a float containing a magnetic portion that activates a reed switch, and metal-sheathed, ceramic-insulated conductors connecting the reed switch to a monitoring system outside the containment vessel. Special materials and special sealing techniques prevent failure of components and allow the float level switch to be connected to a monitoring system outside the containment vessel. 1 figures

Float level switch for a nuclear power plant containment vessel

Science.gov (United States)

Powell, James G.

1993-01-01

This invention is a float level switch used to sense rise or drop in water level in a containment vessel of a nuclear power plant during a loss of coolant accident. The essential components of the device are a guide tube, a reed switch inside the guide tube, a float containing a magnetic portion that activates a reed switch, and metal-sheathed, ceramic-insulated conductors connecting the reed switch to a monitoring system outside the containment vessel. Special materials and special sealing techniques prevent failure of components and allow the float level switch to be connected to a monitoring system outside the containment vessel.

Capacity of Prestressed Concrete Containment Vessels with Prestressing Loss

International Nuclear Information System (INIS)

SMITH, JEFFREY A.

2001-01-01

Reduced prestressing and degradation of prestressing tendons in concrete containment vessels were investigated using finite element analysis of a typical prestressed containment vessel. The containment was analyzed during a loss of coolant accident (LOCA) with varying levels of prestress loss and with reduced tendon area. It was found that when selected hoop prestressing tendons were completely removed (as if broken) or when the area of selected hoop tendons was reduced, there was a significant impact on the ultimate capacity of the containment vessel. However, when selected hoop prestressing tendons remained, but with complete loss of prestressing, the predicted ultimate capacity was not significantly affected for this specific loss of coolant accident. Concrete cracking occurred at much lower levels for all cases. For cases where selected vertical tendons were analyzed with reduced prestressing or degradation of the tendons, there also was not a significant impact on the ultimate load carrying capacity for the specific accident analyzed. For other loading scenarios (such as seismic loading) the loss of hoop prestressing with the tendons remaining could be more significant on the ultimate capacity of the containment vessel than found for the accident analyzed. A combination of loss of prestressing and degradation of the vertical tendons could also be more critical during other loading scenarios

Development of prestressed concrete containment vessels

International Nuclear Information System (INIS)

Yuji, Hideo; Kuniyoshi, Mutsumu; Nagata, Kaoru

1983-01-01

This paper presents a summary of evaluations for the selection of the structural and prestressing system type to be employed for the first domestic Prestressed Concrete Containment Vessel (PCCV) in Japan. This paper also discusses characteristic features in the design of the liner plate system provided on the PCCV inner surface to assure its leak-tight integrity. Prestressed concrete containment vessels so far constructed in foreign countries are to a considerable extent of different structural types, depending on differences in dome shapes, prestressing systems and number of buttresses. These differences are caused not only by differences in design philosophy and construction practices, but also by difference in the level of technology of the times when the individual containment vessels are being constructed. In the investigation reported herein, the most suitable types of PCCV and Prestressing Systems were determined as the results of an overall comparative evaluation of data and information obtained from PCCV's so far constructed from the design, construction and cost aspects, taking into consideration the seismic criteria, available technology, construction practices, regulations and technical standards in Japan. The function of the liner plate system requires the liner to have enough deformability so that the liner deformation can be consistent with the PCCV concrete deformation. Therefore, in the design of the liner plate system a method for evaluating liner deformability was employed, instead of the stress evaluation method which is widely used in the design of ordinary structures. (author)

Scenario based optimization of a container vessel with respect to its projected operating conditions

Science.gov (United States)

Wagner, Jonas; Binkowski, Eva; Bronsart, Robert

2014-06-01

In this paper the scenario based optimization of the bulbous bow of the KRISO Container Ship (KCS) is presented. The optimization of the parametrically modeled vessel is based on a statistically developed operational profile generated from noon-to-noon reports of a comparable 3600 TEU container vessel and specific development functions representing the growth of global economy during the vessels service time. In order to consider uncertainties, statistical fluctuations are added. An analysis of these data lead to a number of most probable upcoming operating conditions (OC) the vessel will stay in the future. According to their respective likeliness an objective function for the evaluation of the optimal design variant of the vessel is derived and implemented within the parametrical optimization workbench FRIENDSHIP Framework. In the following this evaluation is done with respect to vessel's calculated effective power based on the usage of potential flow code. The evaluation shows, that the usage of scenarios within the optimization process has a strong influence on the hull form.

Pressurization of Containment Vessels from Plutonium Oxide Contents

International Nuclear Information System (INIS)

Hensel, S.

2012-01-01

Transportation and storage of plutonium oxide is typically done using a convenience container to hold the oxide powder which is then placed inside a containment vessel. Intermediate containers which act as uncredited confinement barriers may also be used. The containment vessel is subject to an internal pressure due to several sources including; (1) plutonium oxide provides a heat source which raises the temperature of the gas space, (2) helium generation due to alpha decay of the plutonium, (3) hydrogen generation due to radiolysis of the water which has been adsorbed onto the plutonium oxide, and (4) degradation of plastic bags which may be used to bag out the convenience can from a glove box. The contributions of these sources are evaluated in a reasonably conservative manner.

The design, fabrication, and testing of WETF high-quality, long-term-storage, secondary containment vessels

International Nuclear Information System (INIS)

Fisher, Kane J.

2000-01-01

Los Alamos National Laboratory's Weapons Engineering Tritium Facility (WETF) requires secondary containment vessels to store primary tritium containment vessels. The primary containment vessel provides the first boundary for tritium containment. The primary containment vessel is stored within a secondary containment vessel that provides the secondary boundary for tritium containment. WETF requires high-quality, long-term-storage, secondary tritium containment vessels that fit within a Mound-designed calorimeter. In order to qualify the WETF high-quality, long-term-storage, secondary containment vessels for use at WETF, steps have been taken to ensure the appropriate design, adequate testing, quality in fabrication, and acceptable documentation

Pretest round robin analysis of 1:4-scale prestressed concrete containment vessel model

International Nuclear Information System (INIS)

Hessheimer, M.F.; Luk, V.K.; Klamerus, E.W.; Shibata, S.; Mitsugi, S.; Costello, J.F.

2001-01-01

The work reported herein represents, arguably, the state of the art in the numerical simulation of the response of a prestressed concrete containment vessel (PCCV) model to pressure loads up to failure. A significant expenditure of time and money on the part of the sponsors, contractors, and Round Robin participants was required to meet the objectives. While it is difficult to summarize the results of this extraordinary effort in a few paragraphs, the following observations are offered for the reader's consideration: almost half the participants used ABAQUS as the primary computational tool for performing the pretest analyses. The other participants used a variety of codes, most of which were developed ''in house''. (author)

Elements of thought on corium containment strategy in reactor vessel

International Nuclear Information System (INIS)

2015-01-01

As accidents with core fusion are taken into account for the design of third-generation nuclear reactors, this brief document presents the corium containment strategy for a reactor vessel, its limitations, as well as research programs undertaken by the IRSN in this field. The report describes the controlled management of a severe accident, the major objective being to minimise releases in the environment, that which requires to maintain the reactor containment enclosure tightness. Practical actions are briefly indicated. Key points indicating the feasibility of a strategy of containment in vessel are discussed. The impact of reactor power on the robustness of an approach with containment in vessel is also discussed. An overview of technological evolutions and contributions of researches made by the IRSN is finally proposed

Aseismic safety analysis of a prestressed concrete containment vessel for CPR1000 nuclear power plant

Science.gov (United States)

Yi, Ping; Wang, Qingkang; Kong, Xianjing

2017-01-01

The containment vessel of a nuclear power plant is the last barrier to prevent nuclear reactor radiation. Aseismic safety analysis is the key to appropriate containment vessel design. A prestressed concrete containment vessel (PCCV) model with a semi-infinite elastic foundation and practical arrangement of tendons has been established to analyze the aseismic ability of the CPR1000 PCCV structure under seismic loads and internal pressure. A method to model the prestressing tendon and its interaction with concrete was proposed and the axial force of the prestressing tendons showed that the simulation was reasonable and accurate. The numerical results show that for the concrete structure, the location of the cylinder wall bottom around the equipment hatch and near the ring beam are critical locations with large principal stress. The concrete cracks occurred at the bottom of the PCCV cylinder wall under the peak earthquake motion of 0.50 g, however the PCCV was still basically in an elastic state. Furthermore, the concrete cracks occurred around the equipment hatch under the design internal pressure of 0.4MPa, but the steel liner was still in the elastic stage and its leak-proof function soundness was verified. The results provide the basis for analysis and design of containment vessels.

System for cooling the containment vessel of a nuclear reactor

International Nuclear Information System (INIS)

Costes, Didier.

1982-01-01

The invention concerns a post-accidental cooling system for a nuclear reactor containment vessel. This system includes in series a turbine fed by the moist air contained in the vessel, a condenser in which the air is dried and cooled, a compressor actuated by the turbine and a cooling exchanger. The cold water flowing through the condenser and in the exchanger is taken from a tank outside the vessel and injected by a pump actuated by the turbine. The application is for nuclear reactors under pressure [fr

Scenario based optimization of a container vessel with respect to its projected operating conditions

Directory of Open Access Journals (Sweden)

Jonas Wagner

2014-06-01

Full Text Available In this paper the scenario based optimization of the bulbous bow of the KRISO Container Ship (KCS is presented. The optimization of the parametrically modeled vessel is based on a statistically developed operational profile generated from noon-to-noon reports of a comparable 3600 TEU container vessel and specific development functions representing the growth of global economy during the vessels service time. In order to consider uncertainties, statistical fluctuations are added. An analysis of these data lead to a number of most probable upcoming operating conditions (OC the vessel will stay in the future. According to their respective likeliness an objective function for the evaluation of the optimal design variant of the vessel is derived and implemented within the parametrical optimization workbench FRIENDSHIP Framework. In the following this evaluation is done with respect to vessel's calculated effective power based on the usage of potential flow code. The evaluation shows, that the usage of scenarios within the optimization process has a strong influence on the hull form.

Containment vessel for a nuclear reactor

International Nuclear Information System (INIS)

Yamanari, Sh.; Horiuchi, T.; Sugisaki, T.; Tominaga, K.

1985-01-01

A containment vessel for a nuclear reactor having a dry well for mounting therein a pressure vessel for containing the nuclear reactor, a pressure suppressing chamber having a pool of coolant therein, and a vent pipe device for releasing therethrough into the pool of coolant within the pressure suppressing chamber steam which will be produced as a result of the occurrence of an accident and escape into the dry well. The vent pipe device includes a plurality of vent pipe members inserted in the pool of coolant within the pressure suppressing chamber and each having at least one exhaust port opening in the coolant. The vent pipe members are divided into a plurality of groups in such a manner that the vent pipe members of different groups differ from one another in the length of submerged portions of the vent pipe members interposed between the liquid of the coolant within the pressure suppressing chamber and the exhaust ports of the vent pipe members

Leakage detecting method and device for water tight vessel of wet-type container apparatus

International Nuclear Information System (INIS)

Tanaka, Yoshimi.

1995-01-01

The present invention provides a method of and a device for detecting leakage of a water tight vessel of a wet-type container apparatus for containing a reactor pressure vessel while immersing it water in a reactor container. Namely, in the wet-type container apparatus, the periphery of the pressure vessel is coated with a heat insulation material and the periphery of the heat insulation material is coated with a water tight vessel. The water tight vessel is immersed under water in the reactor container. As a method of detecting leakage of the wet-type container apparatus, gases mixed with helium are supplied into the water tight vessel at a pressure higher than the inner pressure of the reactor container at a lowest position of the reactor pressure vessel. A water level in the reactor container is determined so as to form a space at the top portion of the inside of the reactor container. The helium at the top portion is detected to monitor the leakage of the water tight vessel. With such procedures, even if the water tight vessel is ruptured at any position, helium mixed to the gases is released to water in the reactor container and rise up to the top space and detected by a helium leakage detection device. (I.S.)

EDS V25 containment vessel explosive qualification test report.

Energy Technology Data Exchange (ETDEWEB)

Rudolphi, John Joseph

2012-04-01

The V25 containment vessel was procured by the Project Manager, Non-Stockpile Chemical Materiel (PMNSCM) as a replacement vessel for use on the P2 Explosive Destruction Systems. It is the first EDS vessel to be fabricated under Code Case 2564 of the ASME Boiler and Pressure Vessel Code, which provides rules for the design of impulsively loaded vessels. The explosive rating for the vessel based on the Code Case is nine (9) pounds TNT-equivalent for up to 637 detonations. This limit is an increase from the 4.8 pounds TNT-equivalency rating for previous vessels. This report describes the explosive qualification tests that were performed in the vessel as part of the process for qualifying the vessel for explosive use. The tests consisted of a 11.25 pound TNT equivalent bare charge detonation followed by a 9 pound TNT equivalent detonation.

Proving Test on the Reliability for Reactor Containment Vessel

International Nuclear Information System (INIS)

Takumi, K.; Nonaka, A.

1988-01-01

NUPEC (Nuclear Power Engineering Test Center) has started an eight-year project of Proving Test on the Reliability for Reactor Containment Vessel since June 1987. The objective of this project is to confirm the integrity of containment vessels under severe accident conditions. This paper shows the outline of this project. The test Items are (1) Hydrogen mixing and distribution test, (2) Hydrogen burning test, (3) Iodine trapping characteristics test, and (4) Structural behavior test. Based on the test results, computer codes are verified and as the results of analysis and evaluation by the computer codes, containment integrity is to be confirmed

Posttest analysis of a 1:4-scale prestressed concrete containment vessel model

International Nuclear Information System (INIS)

Dameron, R.A.; Rashid, Y.R.; Hessheimer, M.F.

2003-01-01

The Nuclear Power Engineering Corporation (NUPEC) of Japan and the U.S. Nuclear Regulatory Commission (NRC), Office of Nuclear Regulatory Research, co-sponsored a Cooperative Containment Research Program at Sandia National Laboratories (SNL) in Albuquerque, New Mexico. As part of the program, a prestressed concrete containment vessel (PCCV) model was subjected to a series of overpressurization tests at SNL beginning in July 2000 and culminating in a functional failure mode or Limit State Test (LST) in September 2000 and a Structural Failure Mode Test (SFMT) in November 2001. The PCCV model, uniformly scaled at 1:4, is representative of the containment structure of an actual Pressurized Water Reactor (PWR) plant (OHI-3) in Japan. The objectives of the pressurization tests were to obtain measurement of the structural response to pressure loading beyond design basis accident in order to validate analytical modeling, to find pressure capacity of the model, and to observe its failure mechanisms. This paper compares results of pretest analytical studies of the PCCV model to the PCCV high pressure test measurements and describes results of post-test analytical studies. These analyses have been performed by ANATECH Corp. under contract with Sandia National Laboratories. The post-test analysis represents the third phase of a comprehensive PCCV analysis effort. The first phase consisted of preliminary analyses to determine what finite element models would be necessary for the pretest prediction analyses, and the second phase consisted of the pretest prediction analyses. The principal objectives of the post-test analyses were: (1) to provide insights to improve the analytical methods for predicting the structural response and failure modes of a prestressed concrete containment, and (2) to evaluate by analysis any phenomena or failure mode observed during the test that had not been explicitly predicted by analysis. In addition to summarizing comparisons between measured

Containment vessel stability analysis

International Nuclear Information System (INIS)

Harstead, G.A.; Morris, N.F.; Unsal, A.I.

1983-01-01

The stability analysis for a steel containment shell is presented herein. The containment is a freestanding shell consisting of a vertical cylinder with a hemispherical dome. It is stiffened by large ring stiffeners and relatively small longitudinal stiffeners. The containment vessel is subjected to both static and dynamic loads which can cause buckling. These loads must be combined prior to their use in a stability analysis. The buckling loads were computed with the aid of the ASME Code case N-284 used in conjunction with general purpose computer codes and in-house programs. The equations contained in the Code case were used to compute the knockdown factors due to shell imperfections. After these knockdown factors were applied to the critical stress states determined by freezing the maximum dynamic stresses and combining them with other static stresses, a linear bifurcation analysis was carried out with the aid of the BOSOR4 program. Since the containment shell contained large penetrations, the Code case had to be supplemented by a local buckling analysis of the shell area surrounding the largest penetration. This analysis was carried out with the aid of the NASTRAN program. Although the factor of safety against buckling obtained in this analysis was satisfactory, it is claimed that the use of the Code case knockdown factors are unduly conservative when applied to the analysis of buckling around penetrations. (orig.)

Scenario based optimization of a container vessel with respect to its projected operating conditions

Directory of Open Access Journals (Sweden)

Wagner Jonas

2014-06-01

Full Text Available In this paper the scenario based optimization of the bulbous bow of the KRISO Container Ship (KCS is presented. The optimization of the parametrically modeled vessel is based on a statistically developed operational profile generated from noon-to-noon reports of a comparable 3600 TEU container vessel and specific development functions representing the growth of global economy during the vessels service time. In order to consider uncertainties, statistical fluctuations are added. An analysis of these data lead to a number of most probable upcoming operating conditions (OC the vessel will stay in the future. According to their respective likeliness an objective function for the evaluation of the optimal design variant of the vessel is derived and implemented within the parametrical optimization workbench FRIENDSHIP Framework. In the following this evaluation is done with respect to vessel’s calculated effective power based on the usage of potential flow code. The evaluation shows, that the usage of scenarios within the optimization process has a strong influence on the hull form.

Seismic analysis of a reinforced concrete containment vessel model

International Nuclear Information System (INIS)

Randy, James J.; Cherry, Jeffery L.; Rashid, Yusef R.; Chokshi, Nilesh

2000-01-01

Pre-and post-test analytical predictions of the dynamic behavior of a 1:10 scale model Reinforced Concrete Containment Vessel are presented. This model, designed and constructed by the Nuclear Power Engineering Corp., was subjected to seismic simulation tests using the high-performance shaking table at the Tadotsu Engineering Laboratory in Japan. A group of tests representing design-level and beyond-design-level ground motions were first conducted to verify design safety margins. These were followed by a series of tests in which progressively larger base motions were applied until structural failure was induced. The analysis was performed by ANATECH Corp. and Sandia National Laboratories for the US Nuclear Regulatory Commission, employing state-of-the-art finite-element software specifically developed for concrete structures. Three-dimensional time-history analyses were performed, first as pre-test blind predictions to evaluate the general capabilities of the analytical methods, and second as post-test validation of the methods and interpretation of the test result. The input data consisted of acceleration time histories for the horizontal, vertical and rotational (rocking) components, as measured by accelerometers mounted on the structure's basemat. The response data consisted of acceleration and displacement records for various points on the structure, as well as time-history records of strain gages mounted on the reinforcement. This paper reports on work in progress and presents pre-test predictions and post-test comparisons to measured data for tests simulating maximum design basis and extreme design basis earthquakes. The pre-test analyses predict the failure earthquake of the test structure to have an energy level in the range of four to five times the energy level of the safe shutdown earthquake. The post-test calculations completed so far show good agreement with measured data

Reactor containment vessel

International Nuclear Information System (INIS)

Ochiai, Kanehiro; Hayagumo, Sunao; Morikawa, Matsuo.

1981-01-01

Purpose: To safety and simplify the structure in a reactor containment vessel. Constitution: Steam flow channels with steam jetting ports communicating to coolants are provided between a communication channel and coolants in a pressure suppression chamber. Upon loss of coolant accidents, pressure in a dry well will increase, then force downwards water in an annulus portion and further flow out the water through steam jetting ports into a suppression pool. Thus, the steam flow channel is filled with steams or airs present in the dry well, which are released through the steam jetting ports into the pressure suppression chamber. Even though water is violently vibrated owing to the upward movement of air bubbles and condensation of steam bubbles, the annular portion and the steam jetting ports are filled with steams or the like, direct dynamic loads onto the structures such as communication channels can be avoided. (J.P.N.)

Modeling irradiation embrittlement in reactor pressure vessel steels

International Nuclear Information System (INIS)

Odette, G.R.

1998-01-01

As a result of the popularity of the Agencies report 'Neutron Irradiation Embrittlement of Reactor Pressure Vessel Steels' of 1975, it was decided that another report on this broad subject would be of use. In this report, background and contemporary views on specially identified areas of the subject are considered as self-contained chapters, written by experts. In chapter 10, numerical modeling of irradiation embrittlement in reactor vessel steels are introduced. Physically-based models are developed and their role in advancing the state-of-the-art of predicting irradiation embrittlement of RPV steels is stressed

Limit analysis and design of containment vessels

International Nuclear Information System (INIS)

Save, M.

1984-01-01

In the introduction, the theory of plastic analysis of shells is briefly recalled. Minimum-volume design for assigned load factor at plastic collapse is then considered and optimality criteria are derived for plates and shells of continuously varying or piecewise-constant thickness. In the first part, containers made of metal are examined. Analytical and numerical limit analysis solutions and corresponding experimental results are considered for various types of vessels, including intersecting shells. Attention is given to experimental post-yield behavior. Some tests up to fracture are discussed. New theoretical and experimental results of limit analysis of stiffened cylindrical vessels are presented, in which reinforcing rings are treated as discrete structural element (no smearing out) and due account is taken of their strong curvature. Cases of collapse by instability under internal pressure are pointed out. Minimum-volume design of circular plates and cylindrical shells is then formulated and various examples are presented of sandwich and solid metal structures. Containers of piecewise-constant thickness are given particular attention. Available experimental evidence on minimum-volume design of plates and shells is reviewed and commented upon. The second part deals with reinforced concrete vessels. Cylindrical containers are studied, from both points of view of limit analysis and of limit design with minimum volume of reinforcement. The practical use of the latter solutions is discussed. A third part reviews other loading cases (including cyclic and impact loads) and gives indications on corresponding theories, formulations and solution methods. The last part is devoted to a discussion of the limitations of the methods presented, within the frame of the 'limit states' design philosophy, which is first briefly recalled. Considerations on further research in the field conclude the paper. (orig.)

Storage vessel for containing radiation contaminated material

International Nuclear Information System (INIS)

Ogawa, Kazuya.

1995-01-01

A container pipe and an outer pipe are coaxially assembled integrally in a state where securing spacers are disposed between the container pipe and the outer pipe, and an annular flow channel is formed around the container pipe. Radiation contaminated material-containing body (glass solidified package) is contained in the container pipe. The container pipe and the outer pipe in an integrated state are suspended from a ceiling plug of a cell chamber of a storage vessel, and supporting devices are assembled between the pipes and a support structure. A shear/lug mechanism is used for the supporting devices. The combination of the shear/lug allows radial and vertical movement but restrict horizontal movement of the outer tube. The supporting devices are assembled while visually recognizing the state of the shear/lug mechanism between the outer pipe and the support mechanism. Accordingly, operationability upon assembling the container pipe and the outer pipe is improved. (I.N.)

Method of detecting water leakage in radioactive waste containing vessel

International Nuclear Information System (INIS)

Ishioka, Hitoshi; Takao, Yoshiaki; Hayakawa, Kiyoshige.

1989-01-01

Lower level radioactive wastes formed upon operation of nuclear facilities are processed by underground storage. In this case, a plurality of drum cans packed with radioactive wastes are contained in a vessel and a water soluble dye material is placed at the inside of the vessel. The method of placing the water soluble dye material at the inside of the vessel includes a method of coating the material on the inner surface of the vessel and a method of mixing the material in sands to be filled between each of the drum cans. Then, leakage of water soluble dye material is detected when water intruding from the outside into the vessel is again leached out of the vessel, to detect the water leakage from the inside of the vessel. In this way, it is possible to find a water-invaded vessel before corrosion of the drum can by water intruded into the vessel and leakage of nuclides in the drum can. Accordingly, it is possible to apply treatment such as repair before occurrence of accident and can maintain the safety of radioactive water processing facilities. (I.S.)

Dynamic testing of MFTF containment-vessel structural system

International Nuclear Information System (INIS)

Weaver, H.J.; McCallen, D.B.; Eli, M.W.

1982-01-01

Dynamic (modal) testing was performed on the Magnetic Fusion Test Facility (MFTF) containment vessel. The seismic design of this vessel was heavily dependent upon the value of structural damping used in the analysis. Typically for welded steel vessels, a value of 2 to 3% of critical is used. However, due to the large mass of the vessel and magnet supported inside, we felt that the interaction between the structure and its foundation would be enhanced. This would result in a larger value of damping because vibrational energy in the structure would be transferred through the foundation into the surrounding soil. The dynamic test performed on this structure (with the magnet in place) confirmed this later theory and resulted in damping values of approximately 4 to 5% for the whole body modes. This report presents a brief description of dynamic testing emphasizing the specific test procedure used on the MFTF-A system. It also presents an interpretation of the damping mechanisms observed (material and geometric) based upon the spatial characteristics of the modal parameters

Radioactivity concentration measuring device for radiation waste containing vessel

International Nuclear Information System (INIS)

Goto, Tetsuo.

1994-01-01

The device of the present invention can precisely and accurately measure a radioactive concentration of radioactive wastes irrespective of the radioactivity concentration distribution. Namely, a Ge detector having a collimator and a plurality of radiation detectors are placed at the outside of the radioactive waste containing vessel in such a way that it can rotate and move vertically relative to the vessel. The plurality of radiation detectors detect radiation coefficient signals at an assumed segment unit of a predetermined length in vertical direction and for every predetermined angle unit in the rotational direction. A weight measuring device determines the weight of the vessel. A computer calculates an average density of radioactivity for the region filled with radioactivity based on the determined net weight and radiation coefficient signals assuming that the volume of the radioactivity is constant. In addition, the computer calculates the amount of radioactivity in the assumed segment by conducting γ -ray absorption compensation calculation for the material in the vessel. Each of the amount of radioactivity is integrated to determine the amount of radioactivity in the vessel. (I.S.)

Stress analysis of LOFT containment vessel attachments for the mainsteam and feedwater piping support structures

International Nuclear Information System (INIS)

Finicle, D.P.

1977-01-01

The LOFT Containment Vessel attachments for the Mainsteam and Feedwater Piping Support Structures have been analyzed for operating and faulted loading conditions. This report contains the analysis of the connections to the containment vessel for the most current design and loading. Also contained in this report is the analysis of the piping supports

Seismic analysis of a containment vessel

International Nuclear Information System (INIS)

Toledo, E.M.; Jospin, R.J.; Loula, A.F.D.

1987-01-01

A seismic analysis of a nuclear power plant containment vessel is presented. Usual loads in this kind of analysis like SSE, DBE and SSB loadings are considered. With the response spectra, previously obtained, for the above mentioned loadings one uses the response spectrum techniques in order to obtain estimatives for the maximum values of the stresses. Some considerations about the problem and the approcah used herein, are initially described. Next, the analysed structure geometry and some results, compared with those obtained by using computer code ANSYS are shown. (Author) [pt

Containment performance evaluation of prestressed concrete containment vessels with fiber reinforcement

Energy Technology Data Exchange (ETDEWEB)

Choun, Young Sun; Park, Hyung Kui [Integrated Safety Assessment Division, Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2015-12-15

Fibers in concrete resist the growth of cracks and enhance the postcracking behavior of structures. The addition of fibers into a conventional reinforced concrete can improve the structural and functional performance of safety-related concrete structures in nuclear power plants. The influence of fibers on the ultimate internal pressure capacity of a prestressed concrete containment vessel (PCCV) was investigated through a comparison of the ultimate pressure capacities between conventional and fiber-reinforced PCCVs. Steel and polyamide fibers were used. The tension behaviors of conventional concrete and fiber-reinforced concrete specimens were investigated through uniaxial tension tests and their tension-stiffening models were obtained. For a PCCV reinforced with 1% volume hooked-end steel fiber, the ultimate pressure capacity increased by approximately 12% in comparison with that for a conventional PCCV. For a PCCV reinforced with 1.5% volume polyamide fiber, an increase of approximately 3% was estimated for the ultimate pressure capacity. The ultimate pressure capacity can be greatly improved by introducing steel and polyamide fibers in a conventional reinforced concrete. Steel fibers are more effective at enhancing the containment performance of a PCCV than polyamide fibers. The fiber reinforcement was shown to be more effective at a high pressure loading and a low prestress level.

Containment performance evaluation of prestressed concrete containment vessels with fiber reinforcement

International Nuclear Information System (INIS)

Choun, Young Sun; Park, Hyung Kui

2015-01-01

Fibers in concrete resist the growth of cracks and enhance the postcracking behavior of structures. The addition of fibers into a conventional reinforced concrete can improve the structural and functional performance of safety-related concrete structures in nuclear power plants. The influence of fibers on the ultimate internal pressure capacity of a prestressed concrete containment vessel (PCCV) was investigated through a comparison of the ultimate pressure capacities between conventional and fiber-reinforced PCCVs. Steel and polyamide fibers were used. The tension behaviors of conventional concrete and fiber-reinforced concrete specimens were investigated through uniaxial tension tests and their tension-stiffening models were obtained. For a PCCV reinforced with 1% volume hooked-end steel fiber, the ultimate pressure capacity increased by approximately 12% in comparison with that for a conventional PCCV. For a PCCV reinforced with 1.5% volume polyamide fiber, an increase of approximately 3% was estimated for the ultimate pressure capacity. The ultimate pressure capacity can be greatly improved by introducing steel and polyamide fibers in a conventional reinforced concrete. Steel fibers are more effective at enhancing the containment performance of a PCCV than polyamide fibers. The fiber reinforcement was shown to be more effective at a high pressure loading and a low prestress level

Three dimensional non-linear cracking analysis of prestressed concrete containment vessel

International Nuclear Information System (INIS)

Al-Obaid, Y.F.

2001-01-01

The paper gives full development of three-dimensional cracking matrices. These matrices are simulated in three-dimensional non-linear finite element analysis adopted for concrete containment vessels. The analysis includes a combination of conventional steel, the steel line r and prestressing tendons and the anisotropic stress-relations for concrete and concrete aggregate interlocking. The analysis is then extended and is linked to cracking analysis within the global finite element program OBAID. The analytical results compare well with those available from a model test. (author)

Heissdampfreaktor (HDR) steel-containment-vessel and floodwater-storage-tank structural-dynamics tests

International Nuclear Information System (INIS)

Arendts, J.G.

1982-01-01

Inertance (vibration) testing of two significant vessels at the Heissdampfreaktor (HDR) facility, located near Kahl, West Germany, was recently completed. Transfer functions were obtained for determination of the modal properties (frequencies, mode shapes and damping) of the vessels using two different test methods for comparative purposes. One of the vessels tested was the steel containment vessel (SCV). The SCV is approximately 180 feet high and 65 feet in diameter with a 1.2-inch wall thickness. The other vessel, called the floodwater storage tank (FWST), is a vertically standing vessel approximately 40 feet high and 10 feet in diameter with a 1/2-inch wall thickness. The FWST support skirt is square (in plan views) with its corners intersecting the ellipsoidal bottom head near the knuckle region

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

Nonlinear analysis of pre-stressed concrete containment vessel (PCCV) using the damage plasticity model

Energy Technology Data Exchange (ETDEWEB)

Shokoohfar, Ahmad; Rahai, Alireza, E-mail: rahai@aut.ac.ir

2016-03-15

Highlights: • This paper describes nonlinear analyses of a 1:4 scale model of a (PCCV). • Coupled temp-disp. analysis and concrete damage plasticity are considered. • Temperature has limited effects on correct failure mode estimation. • Higher pre-stressing forces have limited effects on ultimate radial displacements. • Anchorage details of liner plates leads to prediction of correct failure mode. - Abstract: This paper describes the nonlinear analyses of a 1:4 scale model of a pre-stressed concrete containment vessel (PCCV). The analyses are performed under pressure and high temperature effects with considering anchorage details of liner plate. The temperature-time history of the model test is considered as an input boundary condition in the coupled temp-displacement analysis. The constitutive model developed by Chang and Mander (1994) is adopted in the model as the basis for the concrete stress–strain relation. To trace the crack pattern of the PCCV concrete faces, the concrete damage plasticity model is applied. This study includes the results of the thermal and mechanical behaviors of the PCCV subject to temperature loading and internal pressure at the same time. The test results are compared with the analysis results. The analysis results show that the temperature has little impact on the ultimate pressure capacity of the PCCV. To simulate the exact failure mode of the PCCV, the anchorage details of the liner plates around openings should be maintained in the analytical models. Also the failure mode of the PCCV structure hasn’t influenced by hoop tendons pre-stressing force variations.

Applicability of JIS SPV 50 steel to primary containment vessel of nuclear power station

International Nuclear Information System (INIS)

Iida, Kunihiro; Ishikawa, Koji; Sakai, Keiichi; Onozuka, Masakazu; Sato, Makoto.

1979-01-01

The space within reactor containment vessels must be expanded in order to improve the reliability of nuclear power plants, accordingly the adoption of large reactor containment vessels is investigated. SGV 42 and 49 steels in JIS G 3118 have been used for containment vessels so far, but stress relief annealing is required when the thickness exceeds 38 mm. The time has come when the use of thicker conventional plates without stress relieving or the use of high strength steel must be examined in detail. In this study, the tests of confirming material properties were carried out on SPV 50 in JIS G 3115, Steels for pressure vessels, aiming at the method of fabrication without stress relieving. The highest and lowest temperatures in use were set at 171 deg and -8 deg C, respectively. The chemical composition and the mechanical properties of the plates tested, the method of welding, the results of tensile test on the parent metal and the welds, the required lowest preheating temperature, the fracture toughness at low temperature and the brittle fracture causing test are reported. The parent metal and the welded joints of SPV 50 have the properties suitable to reactor containment vessels, namely the sufficient fracture toughness to guarantee the prevention of unstable fracture when the method of welding without stress relieving is adopted. (Kako, I.)

Buckling of steel containment shells. Task 1b. Buckling of Washington Public Power Supply Systems' plant No. 2 containment vessel. Final report, 25 August 1980-30 September 1982

International Nuclear Information System (INIS)

Meller, E.; Bushnell, D.

1982-12-01

Static buckling analyses of the steel containment vessel of the Washington Public Power Supply Systems' (WPPSS) plant No. 2 were conducted with use of several computer programs developed at the Lockheed Missiles and Space Company (LMSC). These analyses were conducted as part of Task 1, Evaluation of Two Steel Containment Designs. The report is divided into two main sections. The first gives results from analyses of the containment as if it were axisymmetric (computerized models with use of BOSOR4, BOSOR5, and PANDA), and the second gives results from a STAGSC-1 model in which the largest penetration is included. Good agreement is obtained from analyses with BOSOR5 and STAGSC-1 for a case in which both of these computer programs were applied to the same configuration and loading. It is important to include nonlinear material behavior (plasticity) in the computerized models for collapse. Predictions of collapse from STAGSC-1 indicate that the largest penetration of the WPPSS-2 containment vessel is reinforced such that there is no decrease in load carrying capability below that indicated from models in which this penetration is neglected

Safety margin evaluation of pre-stressed concrete nuclear containment vessel model with BARC code ULCA

International Nuclear Information System (INIS)

Basha, S.M.; Patnaik, R.; Ramanujam, S.; Singh, R.K.; Kushwaha, H.S.; Venkat Raj, V.

2002-01-01

Full text: Ultimate load capacity assessment of nuclear containments has been a thrust research area for Indian pressurised heavy water reactor (PHWR) power programme. For containment safety assessment of Indian PHWRs a finite element code ULCA was developed at BARC, Trombay. This code has been extensively benchmarked with experimental results and for prediction of safety margins of Indian PHWRs. The present paper highlights the analysis results for prestressed concrete containment vessel (PCCV) tested at Sandia National Labs, USA in a round robin analysis activity co-sponsored by Nuclear Power Engineering Corporation (NUPEC), Japan and the U.S Nuclear Regulatory Commission (NRC). Three levels of failure pressure predictions namely the upper bound, the most probable and the lower bound (all with 90% confidence) were made as per the requirements of the round robin analysis activity. The most likely failure pressure is predicted to be in the range of 2.95 Pd to 3.15 Pd (Pd = design pressure of 0.39 MPa for the PCCV model) depending on the type of liners used in the construction of the PCCV model. The lower bound value of the ultimate pressure of 2.80 Pd and the upper bound of the ultimate pressure of 3.45 Pd are also predicted from the analysis. These limiting values depend on the assumptions of the analysis for simulating the concrete tendon interaction and the strain hardening characteristics of the steel members. The experimental test has been recently concluded at Sandia Laboratory and the peak pressure reached during the test is 3.3 Pd that is enveloped by our upper bound prediction of 3.45 Pd and is close to the predicted most likely pressure of 3.15 Pd

Using An Adapter To Perform The Chalfant-Style Containment Vessel Periodic Maintenance Leak Rate Test

International Nuclear Information System (INIS)

Loftin, B.; Abramczyk, G.; Trapp, D.

2011-01-01

Recently the Packaging Technology and Pressurized Systems (PT and PS) organization at the Savannah River National Laboratory was asked to develop an adapter for performing the leak-rate test of a Chalfant-style containment vessel. The PT and PS organization collaborated with designers at the Department of Energy's Pantex Plant to develop the adapter currently in use for performing the leak-rate testing on the containment vessels. This paper will give the history of leak-rate testing of the Chalfant-style containment vessels, discuss the design concept for the adapter, give an overview of the design, and will present results of the testing done using the adapter.

Application of the ASME code in designing containment vessels for packages used to transport radioactive materials

International Nuclear Information System (INIS)

Raske, D.T.; Wang, Z.

1992-01-01

The primary concern governing the design of shipping packages containing radioactive materials is public safety during transport. When these shipments are within the regulatory jurisdiction of the US Department of Energy, the recommended design criterion for the primary containment vessel is either Section III or Section VIII, Division 1, of the ASME Boiler and Pressure Vessel Code, depending on the activity of the contents. The objective of this paper is to discuss the design of a prototypic containment vessel representative of a packaging for the transport of high-level radioactive material

Review on experiments relating to primary containment vessel failure

International Nuclear Information System (INIS)

Suzuki, Hiroyuki; Okada, Hidetoshi; Uchida, Sunsuke; Naitoh, Masanori

2015-01-01

Experiments regarding failures of primary containment vessels (PCVs) are reviewed and remained issues to be investigated in the future are discussed. Experiments are categorized as those relating to criteria of PCV failures and to FP releases through breaches on PCV boundaries. In the experiments categorized as those relating to criteria of PCV failures, experiments with full-scale, scale models, and compounds used for sealing are surveyed. Experiments relating to an amount of radioactive fission products (FPs) trapped at breaches on PCV boundaries are also reviewed. As remained issues to be investigated in the future, two items are pointed out: Evaluating degradation behavior of PCV boundaries exposed to temperature and pressure from the failure onset criteria to far above them, and evaluating an amount of FPs trapped at breaches on PCV boundaries. (author)

Ductile fracture of cylindrical vessels containing a large flaw

Science.gov (United States)

Erdogan, F.; Irwin, G. R.; Ratwani, M.

1976-01-01

The fracture process in pressurized cylindrical vessels containing a relatively large flaw is considered. The flaw is assumed to be a part-through or through meridional crack. The flaw geometry, the yield behavior of the material, and the internal pressure are assumed to be such that in the neighborhood of the flaw the cylinder wall undergoes large-scale plastic deformations. Thus, the problem falls outside the range of applicability of conventional brittle fracture theories. To study the problem, plasticity considerations are introduced into the shell theory through the assumptions of fully-yielded net ligaments using a plastic strip model. Then a ductile fracture criterion is developed which is based on the concept of net ligament plastic instability. A limited verification is attempted by comparing the theoretical predictions with some existing experimental results.

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

Experimental study of the structural behavior of the reinforced concrete containment vessel beyond design pressure

International Nuclear Information System (INIS)

Oyamada, O.; Saito, H.; Muramatsu, Y.; Hasegawa, T.; Tanaka, N.

1990-01-01

The first Advanced Boiling Water Reactor (ABWR) including a reinforced concrete containment vessel (RCCV) is scheduled to be constructed in the 1990s, in Japan. As the RCCV is new to Japan, we performed a trial design, several series of fundamental experiments and partial/total model experiments. This paper presents a summary of the 'TOP SLAB EXPERIMENT' carried out as one of partial model experiments, in which the structural behavior of the RCCV was examined under internal pressure. (orig.)

Ex-Vessel Core Melt Modeling Comparison between MELTSPREAD-CORQUENCH and MELCOR 2.1

Energy Technology Data Exchange (ETDEWEB)

Robb, Kevin R. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Farmer, Mitchell [Argonne National Lab. (ANL), Argonne, IL (United States); Francis, Matthew W. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

2014-03-01

System-level code analyses by both United States and international researchers predict major core melting, bottom head failure, and corium-concrete interaction for Fukushima Daiichi Unit 1 (1F1). Although system codes such as MELCOR and MAAP are capable of capturing a wide range of accident phenomena, they currently do not contain detailed models for evaluating some ex-vessel core melt behavior. However, specialized codes containing more detailed modeling are available for melt spreading such as MELTSPREAD as well as long-term molten corium-concrete interaction (MCCI) and debris coolability such as CORQUENCH. In a preceding study, Enhanced Ex-Vessel Analysis for Fukushima Daiichi Unit 1: Melt Spreading and Core-Concrete Interaction Analyses with MELTSPREAD and CORQUENCH, the MELTSPREAD-CORQUENCH codes predicted the 1F1 core melt readily cooled in contrast to predictions by MELCOR. The user community has taken notice and is in the process of updating their systems codes; specifically MAAP and MELCOR, to improve and reduce conservatism in their ex-vessel core melt models. This report investigates why the MELCOR v2.1 code, compared to the MELTSPREAD and CORQUENCH 3.03 codes, yield differing predictions of ex-vessel melt progression. To accomplish this, the differences in the treatment of the ex-vessel melt with respect to melt spreading and long-term coolability are examined. The differences in modeling approaches are summarized, and a comparison of example code predictions is provided.

Minimizing Lid Overstows in Master Stowage Plans for Container Vessels is NP-Complete

DEFF Research Database (Denmark)

Ajspur, Mai Lise; Jensen, Rune Møller; Guilbert, Nicolas

Container vessel stowage is a particularly hard combinatorial problem within the shipping industry. The currently most successful approaches decompose the problem hierarchically and first generate a master plan that handle highlevel constraints and objectives such as balance and stress moments...... that it is an NP -complete problem to generate master plans that minimize the number of these lid overstows. Since any efficient approach to container vessel stowage most likely must include a master plan, the implication of this result is that future research must focus and developing good heuristics...

A photoelastic study of the effects of an impulsive seismic wave on a nuclear containment vessel

International Nuclear Information System (INIS)

Burger, C.P.

1981-01-01

A dynamic photoelastic study of the progressive movement of a dilatational P-wave into a model of a nuclear containment vessel,is studied. The reflections at the dome abutments are observed and the strong flexural wave that deforms the dome itself is studied with photoelasticity and with dynamic strain gage procedures. (E.G.) [pt

Welding the AT-400A Containment Vessel

Energy Technology Data Exchange (ETDEWEB)

Brandon, E.

1998-11-01

Early in 1994, the Department of Energy assigned Sandia National Laboratories the responsibility for designing and providing the welding system for the girth weld for the AT-400A containment vessel. (The AT-400A container is employed for the shipment and long-term storage of the nuclear weapon pits being returned from the nation's nuclear arsenal.) Mason Hanger Corporation's Pantex Plant was chosen to be the production facility. The project was successfully completed by providing and implementing a turnkey welding system and qualified welding procedure at the Pantex Plant. The welding system was transferred to Pantex and a pilot lot of 20 AT-400A containers with W48 pits was welded in August 1997. This document is intended to bring together the AT-400A welding system and product (girth weld) requirements and the activities conducted to meet those requirements. This document alone is not a complete compilation of the welding development activities but is meant to be a summary to be used with the applicable references.

Combining endoscopes with PIV and digital holography for the study of vessel model mechanics

International Nuclear Information System (INIS)

Arévalo, Laura; Palero, Virginia; Andrés, Nieves; Arroyo, M P; Lobera, Julia

2015-01-01

In this work traditional fluid and solid mechanics measurement techniques have been combined with endoscopes for the study of blood vessel models’ mechanical properties. Endoscopes have been used as the imaging part of a high-speed PIV system to obtain the velocity field in a vessel model immersed in a container with a refractive index-matching liquid. In this way, we take advantage of the fact that the endoscope tip can be immersed in liquid. Endoscopes have also been used as the imaging and illuminating part of a digital holographic set-up for wall deformation measurement. The novelty of this work is that only one endoscope was used for illuminating and observing the vessel model, using the endoscope’s own illuminating system as the illumination source. The performance of endoscopes in different vessel models has been tested. The results of flow velocity and wall deformation in the different blood vessel models are presented. (paper)

Model tests for prestressed concrete pressure vessels

International Nuclear Information System (INIS)

Stoever, R.

1975-01-01

Investigations with models of reactor pressure vessels are used to check results of three dimensional calculation methods and to predict the behaviour of the prototype. Model tests with 1:50 elastic pressure vessel models and with a 1:5 prestressed concrete pressure vessel are described and experimental results are presented. (orig.) [de

Modelling of hydrogen deflagration in vessels using GOTHIC

International Nuclear Information System (INIS)

Wang, L.L.; Wong, R.C.; Fluke, R.J.

1997-01-01

Simulations of hydrogen deflagration tests were performed using the discrete lumpedparameter bum model of the computer code GOTHIC. The tests were performed in small and large scale spherical vessels and a cylindrical vessel. The small vessel cases included the effects of venting, and the cylindrical tests included the effects of obstacles. The simulations were performed by sub-dividing the volumes into either five or ten 'cells', and parameters such as flame speed and hydrogen concentration were varied. Measured flame speeds were used in the simulations and the results were compared to simulations using the code 'default' flame speed. The calculated pressure transients compared well with the experimental results using the measured flame speeds in the simulations of unvented cases, whereas for vented cases, the predicted peak pressures were generally less than the measurements. However, when the code default flame speed is used, the predicted peak pressures were more consistent and generally conservative when compared with the measurements. When the default flame speeds were used for vessels without obstacles, the peak pressures obtained were higher and the bum times were shorter than the experimental measurements. This was probably due to the basis for the correlations used for default flame speed in the bum model. These correlations were derived from intermediate-scale experiments for hydrogen combustion in relatively turbulent (fans on) environments. For vessels without obstacles, laminar flame speeds were more likely. Hence, the predicted peak pressures would be expected to be higher than the experimental results. In order to account for the degree of turbulence and flame acceleration caused by the presence of obstacles, higher than default flame speeds were used in the simulation of the vessel with obstacles. It was found that twice the default flame speed provided predictions of peak pressures comparable to the measurements. Based on the simulations conducted

Validation of ASTEC core degradation and containment models

International Nuclear Information System (INIS)

Kruse, Philipp; Brähler, Thimo; Koch, Marco K.

2014-01-01

Ruhr-Universitaet Bochum performed in a German funded project validation of in-vessel and containment models of the integral code ASTEC V2, jointly developed by IRSN (France) and GRS (Germany). In this paper selected results of this validation are presented. In the in-vessel part, the main point of interest was the validation of the code capability concerning cladding oxidation and hydrogen generation. The ASTEC calculations of QUENCH experiments QUENCH-03 and QUENCH-11 show satisfactory results, despite of some necessary adjustments in the input deck. Furthermore, the oxidation models based on the Cathcart–Pawel and Urbanic–Heidrick correlations are not suitable for higher temperatures while the ASTEC model BEST-FIT based on the Prater–Courtright approach at high temperature gives reliable enough results. One part of the containment model validation was the assessment of three hydrogen combustion models of ASTEC against the experiment BMC Ix9. The simulation results of these models differ from each other and therefore the quality of the simulations depends on the characteristic of each model. Accordingly, the CPA FRONT model, corresponding to the simplest necessary input parameters, provides the best agreement to the experimental data

Development of containers sealing system like part of surveillance program of the vessel in nuclear power plants

International Nuclear Information System (INIS)

Romero C, J.; Hernandez C, R.; Fernandez T, F.; Rocamontes A, M.; Perez R, N.

2009-10-01

The owners of nuclear power plants should be demonstrate that the embrittlement effects by neutronic radiation do not commit the structural integrity from the pressure vessel of nuclear reactors, during conditions of routine operation and below postulate accident. For this reason, there are surveillance programs of vessels of nuclear power plants, in which are present surveillance capsules. A surveillance capsule is compound by the support, six containers for test tubes and dosimeters. The containers for test tubes are of two types: rectangular container for test tubes, Charpy V and Cylindrical Container for tension test tubes. These test tubes are subject to a same or bigger neutronic flow to that of vessel, being representative of vessel mechanical conditions. The test tubes are rehearsed to watch over the increase of embrittlement that presents the vessel. This work describes the development of welding system to seal the containers for test tubes, these should be filled with helium of ultra high purity, to a pressure of an atmosphere. In this system the welding process Gas Tungsten Arc Welding is used, a hermetic camera that allows to place the containers with three grades of freedom, a vacuum subsystem and pressure, high technology equipment's like: power source with integrated computer, arc starter of high frequency, helium flow controller, among others. Finally, the advances in the inspection system for the qualification of sealing system are mentioned, system that should measure the internal pressure of containers and the helium purity inside these. (Author)

A three-temperature model of selective photothermolysis for laser treatment of port wine stain containing large malformed blood vessels

International Nuclear Information System (INIS)

Li, D.; Wang, G.X.; He, Y.L.; Wu, W.J.; Chen, B.

2014-01-01

As congenital vascular malformations, port wine stain (PWS) is composed of ectatic venular capillary blood vessels buried within healthy dermis. In clinic, pulsed dye laser (PDL) in visible band (e.g. 585 nm) together with cryogen spray cooling (CSC) have become the golden standard for treatment of PWS. However, due to the limited energy deposition of the PDL in blood, large blood vessels are likely to survive from the laser irradiation. As a result, complete clearance of the lesions is rarely achieved. Assuming the local thermal non-equilibrium in skin tissue during the laser surgery, a three-temperature model is proposed to treat the PWS tissue as a porous media composed of a non-absorbing dermal matrix buried with the blood as well as the large malformed blood vessels. Three energy equations are constructed and solved coupling for the temperature of the blood in average-sized PWS vessels, non-absorbing dermal tissues and large malformed blood vessels, respectively. Subsequently, the thermal responses of human skin to visible (585 nm) and near-infrared (1064 nm) laser irradiations with various pulse durations in conjunction with cryogen spray cooling are investigated by the new model, and Arrhenius integral is used to analyze the thermal damage. The simulations show that the short pulse duration of 1.5 ms results in a higher selective heating of blood over epidermis, which will lead to a desired clinic outcome than the longer pulse duration. Due to a much deeper light penetration depth, laser irradiation with 1064 nm in wavelength is superior to that with 585 nm in treating patients with cutaneous hyper-vascular malformation. Complete coagulations are predicted in large-sized and deeply extending blood vessels by 1064 nm laser. - Highlights: •A three-temperature model is proposed for the laser treatment of port wine stain (PWS). •Average sized and large malformed blood vessels in porous medium (tissue) are considered. •Thermal responses of PWS to

Air and gas cleaning methods for reactor containment vessels

Energy Technology Data Exchange (ETDEWEB)

Silverman, L.

1963-11-15

In this paper, a survey is made of the existing and some proposed new methods for the control and purification of air and gases which might be released from a reactor contained or confined for protection of the health and safety of the public from potential accidents. The difference between confinement and containment concepts must be considered. The problems involved and the need for decontamination, site selection, exclusion area, population density, distance, etc., have been discussed elsewhere. We propose to discuss here the safety measures necessary to control the release of radioactive materials to the environment. This requires special systems which must function effectively to minimize loss of fission products such as halogens and particulates. These can penetrate the confinement filters or the containment vessel to a limited extent even after cleaning.

A simple evaluation of containment integrity against ex-vessel steam explosion

International Nuclear Information System (INIS)

Nishiura, Hiroshi

2000-01-01

The guideline for consideration to severe accidents on containment design for next-generation LWR was published in 1999. In order to verify the validity of future containment designs, we have developed a method of assessing for the containment integrity against ex-vessel steam explosion. First, we conducted a simple evaluation on an Advanced PWR. The strength of the reactor cavity wall was assumed to be equivalent to the total strain energy which would accumulate by the time one reinforcing bar element would first reach the failure strain in FEM analyses. As a result, the strength was evaluated to be about 72 MJ. The explosion energy was assumed to be a function of the mass of the dropping melted core and the conversion ratio. Assuming the conversion ratio of 1%, it was estimated that the explosion energy would amount to about 1 MJ if the melt mass corresponds to the break of one instrumentation guide tube penetration, and about 40 MJ if the mass corresponds to the simultaneous break of all penetrations. Therefore, it is expected that the explosion energy would be less than the wall strength; thus, the containment integrity would be maintained even if an ex-vessel steam explosion were to occur. (author)

Modelling of containment atmosphere mixing and stratification experiment using CFD approach

International Nuclear Information System (INIS)

Ivo Kljenak; Miroslav Babic; Borut Mavko; Ivan Bajsic

2005-01-01

An experiment on containment atmosphere mixing and stratification, which was originally performed in the TOSQAN facility in Saclay (France), was simulated with the Computational Fluid Dynamics code CFX. The TOSQAN facility consists of a large cylindrical vessel in which gases are injected. In the considered experiment, steam, air and helium were injected during different phases of the experiment, with steam condensing on vessel walls. Three intermediate steady states, which were obtained with different boundary conditions, were simulated independently. A two-dimensional axisymmetric model of the TOSQAN vessel for the CFX4.4 code was developed. The flow in the simulation domain was modelled as single-phase. Steam condensation on vessel walls was modelled as a sink of mass and energy. Calculated profiles of temperature, steam concentration, and velocity components are compared to experimental results. (authors)

Device for protecting the containment vessel dome of a nuclear reactor

International Nuclear Information System (INIS)

Allain, A.; Filloleau, E.; Mulot, P.

1976-01-01

A device is disclosed for protecting the dome of a nuclear reactor containment vessel against the upward displacement of the concrete shield slab of said reactor and the resultant effects of tilting of an equipment unit mounted on the shield slab at the periphery of said slab, wherein said device comprises: (1) means for separating the equipment unit into two sections consisting of an upper section and a lower section, said lower section being rigidly fixed to said shield slab and said means being actuated by the upward displacement of said slab, (2) a system for vertical rectilinear guiding of said upper section within the containment vessel, and (3) rigid mechanical components which provide a coupling between the aforesaid upper and lower sections of the equipment unit and exert on said upper section under the action of the tilting motion of said lower section a thrust which causes the upward displacement of said upper section

Applicability of JIS SPV 50 steel to primary containment vessels of nuclear power stations

International Nuclear Information System (INIS)

Iida, K.; Ishikawa, K.; Satoh, M.; Soya, I.

1980-01-01

The fracture toughness of JIS SPV 50 steel and its weldment has been examined in order to verify the applicability of these materials to primary containment vessels of nuclear power stations. Test results were evaluated using elastic plastic fracture mechanics through the COD and the J integral concepts for non ductile fracture initiation characteristics. Linear fracture mechanics was employed for propagation arrest characteristics. Results showed that the materials tested here have a sufficient fracture toughness to prevent nonductile fracture and that this steel is a suitable material for use in construction of primary containment vessels of nuclear power stations. (author)

BBRV post-tensioning systems as applied to reactor containments and prestressed concrete pressure vessels

International Nuclear Information System (INIS)

Thorpe, W.; Speck, F.E.

1976-01-01

Nuclear containments and pressure vessels can be post-tensioned by using two basically different methods: tendons and winding. The fundamental differences between the two concepts are shown by introductory examples. A discussion of tendon units, usually lying in the range 4000 to 10,000 kN, is followed by a detailed presentation of the BBRV winding system. After giving a short comment to factors influencing the choice of a post-tensioning system the authors discuss specific aspects of some application groups: cable layout with containments and pressure vessels, conditions for a wrapped design, corrosion protection. (author)

LANL Robotic Vessel Scanning

Energy Technology Data Exchange (ETDEWEB)

Webber, Nels W. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

2015-11-25

Los Alamos National Laboratory in J-1 DARHT Operations Group uses 6ft spherical vessels to contain hazardous materials produced in a hydrodynamic experiment. These contaminated vessels must be analyzed by means of a worker entering the vessel to locate, measure, and document every penetration mark on the vessel. If the worker can be replaced by a highly automated robotic system with a high precision scanner, it will eliminate the risks to the worker and provide management with an accurate 3D model of the vessel presenting the existing damage with the flexibility to manipulate the model for better and more in-depth assessment.The project was successful in meeting the primary goal of installing an automated system which scanned a 6ft vessel with an elapsed time of 45 minutes. This robotic system reduces the total time for the original scope of work by 75 minutes and results in excellent data accumulation and transmission to the 3D model imaging program.

Manufacturing method for radioactive material containing vessel

International Nuclear Information System (INIS)

Kamino, Yoshikazu; Nishioka, Eiji; Toyota, Michinori.

1997-01-01

A containing vessel for radioactive materials (for example, spent fuels) comprises an inner cylinder made of stainless steel having a space for containing radioactive materials at the inside and an outer cylinder made of stainless steel disposed at the outer side of the inner cylinder. Lead homogenization is applied to a space between the inner and the outer cylinders to deposit a lead layer. Then, molten lead heated to a predetermined temperature is cast into the space between the inner and the outer cylinders. A valve is opened to discharge the molten lead in the space from a molten lead discharge pipe, and heated molten lead is injected from a molten lead supply pipe. Then, the discharge of the molten lead and the injection of the molten lead are stopped, and the lead in the space is coagulated. With such procedures, gaps are not formed between the lead of the homogenized portion and the lead of cast portion even when the thickness of the inner and the outer cylinders is great. (I.N.)

Method of detecting leakage in nuclear reactor containment vessel

International Nuclear Information System (INIS)

Koba, Akitoshi; Goto, Seiichiro.

1974-01-01

Object: To permit accurate and prompt detection of leakage of a radioactive substance. Structure: The rate of change of such factors as radiation dose, temperature and pressure in the containment vessel, and each detected rate of change is compared with a reference value. The running cycle of the condensed drain exhausting pump in a drain collecting tank within a predetermined period is detected, and it is also compared with a reference value. These comparisons determine the absence or presence of leakage. (Kamimura, M.)

Network Design Models for Container Shipping

DEFF Research Database (Denmark)

Reinhardt, Line Blander; Kallehauge, Brian; Nielsen, Anders Nørrelund

This paper presents a study of the network design problem in container shipping. The paper combines the network design and fleet assignment problem into a mixed integer linear programming model minimizing the overall cost. The major contributions of this paper is that the time of a vessel route...... is included in the calculation of the capacity and that a inhomogeneous fleet is modeled. The model also includes the cost of transshipment which is one of the major cost for the shipping companies. The concept of pseudo simple routes is introduced to expand the set of feasible routes. The linearization...

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

International Nuclear Information System (INIS)

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

2010-01-01

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

Investigation of radial shear in the wall-base juncture of a 1:4 scale prestressed concrete containment vessel model

Energy Technology Data Exchange (ETDEWEB)

Dameron, R.A.; Rashid, Y.R. [ANATECH Corp., San Diego, CA (United States); Luk, V.K.; Hessheimer, M.F. [Sandia National Labs., Albuquerque, NM (United States)

1998-04-01

Construction of a prestressed concrete containment vessel (PCCV) model is underway as part of a cooperative containment research program at Sandia National Laboratories. The work is co-sponsored by the Nuclear Power Engineering Corporation (NUPEC) of Japan and US Nuclear Regulatory Commission (NRC). Preliminary analyses of the Sandia 1:4 Scale PCCV Model have determined axisymmetric global behavior and have estimated the potential for failure in several areas, including the wall-base juncture and near penetrations. Though the liner tearing failure mode has been emphasized, the assumption of a liner tearing failure mode is largely based on experience with reinforced concrete containments. For the PCCV, the potential for shear failure at or near the liner tearing pressure may be considerable and requires detailed investigation. This paper examines the behavior of the PCCV in the region most susceptible to a radial shear failure, the wall-basemat juncture region. Prediction of shear failure in concrete structures is a difficult goal, both experimentally and analytically. As a structure begins to deform under an applied system of forces that produce shear, other deformation modes such as bending and tension/compression begin to influence the response. Analytically, difficulties lie in characterizing the decrease in shear stiffness and shear stress and in predicting the associated transfer of stress to reinforcement as cracks become wider and more extensive. This paper examines existing methods for representing concrete shear response and existing criteria for predicting shear failure, and it discusses application of these methods and criteria to the study of the 1:4 scale PCCV.

Investigation of radial shear in the wall-base juncture of a 1:4 scale prestressed concrete containment vessel model

International Nuclear Information System (INIS)

Dameron, R.A.; Rashid, Y.R.; Luk, V.K.; Hessheimer, M.F.

1998-04-01

Construction of a prestressed concrete containment vessel (PCCV) model is underway as part of a cooperative containment research program at Sandia National Laboratories. The work is co-sponsored by the Nuclear Power Engineering Corporation (NUPEC) of Japan and US Nuclear Regulatory Commission (NRC). Preliminary analyses of the Sandia 1:4 Scale PCCV Model have determined axisymmetric global behavior and have estimated the potential for failure in several areas, including the wall-base juncture and near penetrations. Though the liner tearing failure mode has been emphasized, the assumption of a liner tearing failure mode is largely based on experience with reinforced concrete containments. For the PCCV, the potential for shear failure at or near the liner tearing pressure may be considerable and requires detailed investigation. This paper examines the behavior of the PCCV in the region most susceptible to a radial shear failure, the wall-basemat juncture region. Prediction of shear failure in concrete structures is a difficult goal, both experimentally and analytically. As a structure begins to deform under an applied system of forces that produce shear, other deformation modes such as bending and tension/compression begin to influence the response. Analytically, difficulties lie in characterizing the decrease in shear stiffness and shear stress and in predicting the associated transfer of stress to reinforcement as cracks become wider and more extensive. This paper examines existing methods for representing concrete shear response and existing criteria for predicting shear failure, and it discusses application of these methods and criteria to the study of the 1:4 scale PCCV

Device for the simultaneous operation of the closing valve of a vessel and the closing valve of a transport container

International Nuclear Information System (INIS)

Tellier, Claude; Surriray, Michel.

1982-01-01

This device includes mechanisms for unlatching the closing valve of the vessel and securing it to the closing valve of the transport container and other mechanisms for vertically raising the assembly of valves, pivoting it and bringing it into a vertical position in a bulge provided in the bottom of the transport container. For example the first containment is a nuclear reactor vessel and the transport container is used for carrying an item from the vessel to an external area (for instance, a defective pump to the repair area) and for the return transport operation [fr

Development of improved SGV480 steel plate for containment vessel in PWR plants

Energy Technology Data Exchange (ETDEWEB)

Murakami, Norioki [Advanced Nuclear Equipment Research Inst., Tokyo (Japan); Morikage, Yasushi; Okayama, Yutaka; Higashikubo, Tomohiro

2001-01-01

When a nuclear containment vessel made of steel plate at PWR plants in Japan is produced, SGV480 steel plate made by annealing method according to JIS G3118 is usually used in main. And, when thickness of welding portion of the vessel is larger than 38 mm, as heat treatment after welding is regulated to carry out according to the ministerial ordinance, it is difficult in actual to carry out the heat treatment of the actual welded portions. In a leading plant, approval of welding using a special method without heat treatment less than 47.25 mm of SGV480 carbon steel plate for JIS G3118 middle and ordinary pressure vessel was carried out to supply it for actual use. And, it is required for protection of welding fracture to carry out pre-heat treatment before welding. Because of increasing plate thickness requiring for lower temperature and more seismic resistance in construction condition, in order to produce a containment vessel without heat treatment after welding, more toughness is required for using material and welded portion. Therefore, a new SGV480 steel plate was developed by using TMCP method of modern steel manufacturing technology, to establish lower carbon equivalence and finer texture with upgrading of both toughness and weldability, without heat treatment after welding and pre-heat treatment before welding, at the Shin-Nippon Steel Co, Ltd. and Kawasaki Steel, Co. Ltd., respectively. (G.K.)

Mouse lung contains endothelial progenitors with high capacity to form blood and lymphatic vessels

Directory of Open Access Journals (Sweden)

Barleon Bernhard

2010-07-01

Full Text Available Abstract Background Postnatal endothelial progenitor cells (EPCs have been successfully isolated from whole bone marrow, blood and the walls of conduit vessels. They can, therefore, be classified into circulating and resident progenitor cells. The differentiation capacity of resident lung endothelial progenitor cells from mouse has not been evaluated. Results In an attempt to isolate differentiated mature endothelial cells from mouse lung we found that the lung contains EPCs with a high vasculogenic capacity and capability of de novo vasculogenesis for blood and lymph vessels. Mouse lung microvascular endothelial cells (MLMVECs were isolated by selection of CD31+ cells. Whereas the majority of the CD31+ cells did not divide, some scattered cells started to proliferate giving rise to large colonies (> 3000 cells/colony. These highly dividing cells possess the capacity to integrate into various types of vessels including blood and lymph vessels unveiling the existence of local microvascular endothelial progenitor cells (LMEPCs in adult mouse lung. EPCs could be amplified > passage 30 and still expressed panendothelial markers as well as the progenitor cell antigens, but not antigens for immune cells and hematopoietic stem cells. A high percentage of these cells are also positive for Lyve1, Prox1, podoplanin and VEGFR-3 indicating that a considerabe fraction of the cells are committed to develop lymphatic endothelium. Clonogenic highly proliferating cells from limiting dilution assays were also bipotent. Combined in vitro and in vivo spheroid and matrigel assays revealed that these EPCs exhibit vasculogenic capacity by forming functional blood and lymph vessels. Conclusion The lung contains large numbers of EPCs that display commitment for both types of vessels, suggesting that lung blood and lymphatic endothelial cells are derived from a single progenitor cell.

Assessment of Ultimate Load Capacity for Pre-Stressed Concrete Containment Vessel Model of PWR Design With BARC Code ULCA

International Nuclear Information System (INIS)

Basha, S.M.; Singh, R.K.; Patnaik, R.; Ramanujam, S.; Kushwaha, H.S.; Venkat Raj, V.

2002-01-01

Ultimate load capacity assessment of nuclear containments has been a thrust research area for Indian Pressurised Heavy Water Reactor (PHWR) power programme. For containment safety assessment of Indian PHWRs a finite element code ULCA was developed at BARC, Trombay. This code has been extensively benchmarked with experimental results. The present paper highlights the analysis results for Prestressed Concrete Containment Vessel (PCCV) tested at Sandia National Labs, USA in a Round Robin analysis activity co-sponsored by Nuclear Power Engineering Corporation (NUPEC), Japan and the U.S Nuclear Regulatory Commission (NRC). Three levels of failure pressure predictions namely the upper bound, the most probable and the lower bound (all with 90% confidence) were made as per the requirements of the round robin analysis activity. The most likely failure pressure is predicted to be in the range of 2.95 Pd to 3.15 Pd (Pd= design pressure of 0.39 MPa for the PCCV model) depending on the type of liners used in the construction of the PCCV model. The lower bound value of the ultimate pressure of 2.80 Pd and the upper bound of the ultimate pressure of 3.45 Pd are also predicted from the analysis. These limiting values depend on the assumptions of the analysis for simulating the concrete-tendon interaction and the strain hardening characteristics of the steel members. The experimental test has been recently concluded at Sandia Laboratory and the peak pressure reached during the test is 3.3 Pd that is enveloped by our upper bound prediction of 3.45 Pd and is close to the predicted most likely pressure of 3.15 Pd. (authors)

Radioactive material-containing vessel and method of manufacturing the same

International Nuclear Information System (INIS)

Kanazawa, Hiroshi; Wada, Katsuyoshi; Ota, Shigeo; Nishioka, Eiji; Okuno, Michinori.

1995-01-01

In a vessel for containing radioactive materials having an outer wall with a structure of interposing a lead layer, as a shielding material between inner and outer cylinders made of steel plates, the inner cylinder and the lead layer are in close contact by way of a thin layer of a lead/tin type soldering material and to such an extent that the boundary layer is not detected by supersonic inspection. In addition, flux is coated to the steel plate, which forms the inner cylinder, on the surface being in contact with the lead layer, then a thin layer of the soldering material such as lead or tin is formed, to cast the lead between the inner and the outer cylinders. Then, since the inner cylinder and the lead layer are thermally joined tightly, heat generated at the inside can effectively be released to the outside, so that it is effective as a high-performance cask for transporting a large amount of radioactive materials such as spent nuclear fuels having high temperature afterheat. In addition, a containing vessel with good contact between the inner cylinder and the lead can be manufactured at a low cost only applying a simple primer treatment on the surface of the inner cylinder in addition to an existent lead casting method. (N.H.)

Integrated leak rate test results of JOYO reactor containment vessel

International Nuclear Information System (INIS)

Tamura, M.; Endo, J.

1982-02-01

Integrated leak rate tests of JOYO after the reactor coolant system had been filled with sodium have been performed two times since 1978 (February 1978 and December 1979). The tests were conducted with the in-containment sodium systems, primary argon cover gas system and air conditioning systems operating. Both the absolute pressure method and the reference chamber method were employed during the test. The results of both tests confirmed the functioning of the containment vessel, and leak rate limits were satisfied. In Addition, the adequancy of the test instrumentation system and the test method was demonstrated. Finally the plant conditions required to maintain reasonable accuracy for the leak rate testing of LMFBR were established. In this paper, the test conditions and the test results are described. (author)

A new model for anisotropic damage in concrete and its application to the prediction of failure of some containment vessel

International Nuclear Information System (INIS)

Badel, P.-B.; Godard, V.; Leblond, J.-B.

2005-01-01

The aim of this paper is to propose a new model for damage in concrete structures which incorporates such complex features as damage anisotropy and asymmetry between tension and compression, while being expressed in a format well suited for numerical applications and involving a limited number of material parameters which can be determined from standard experiments. A crude version of the model involving a single tonsorial internal variable representing damage in tension, and a single material parameter, is presented first. The predictions of this simple model are satisfactory in simple tension, but not so in simple compression. As a remedy, various refinements are then introduced in a second version of the model involving an additional tonsorial or scalar internal variable representing damage in compression, and five additional material parameters. An example of determination of the model parameters using experimental stress-strain curves in simple tension and compression, plus failure envelopes in biaxial tension/compression, is presented next. The model is finally applied to the numerical prediction of the failure of some containment vessel subjected to some large internal pressure, with a comparison with calculations based on a simpler isotropic variant of the model using a single scalar damage variable. The results illustrate the relevance of models incorporating both asymmetry between tension and compression and anisotropy of damage for simulations of industrial concrete structures. (authors)

Special enclosure for a pressure vessel

International Nuclear Information System (INIS)

Wedellsborg, B.W.; Wedellsborg, U.W.

1993-01-01

A pressure vessel enclosure is described comprising a primary pressure vessel, a first pressure vessel containment assembly adapted to enclose said primary pressure vessel and be spaced apart therefrom, a first upper pressure vessel jacket adapted to enclose the upper half of said first pressure vessel containment assembly and be spaced apart therefrom, said upper pressure vessel jacket having an upper rim and a lower rim, each of said rims connected in a slidable relationship to the outer surface of said first pressure vessel containment assembly, mean for connecting in a sealable relationship said upper rim of said first upper pressure vessel jacket to the outer surface of said first pressure vessel containment assembly, means for connecting in a sealable relationship said lower rim of said first upper pressure vessel jacket to the outer surface of said first pressure vessel containment assembly, a first lower pressure vessel jacket adapted to enclose the lower half of said first pressure vessel containment assembly and be spaced apart therefrom, said lower pressure vessel jacket having an upper rim connected in a slidable relationship to the outer surface of said first pressure vessel containment assembly, and means for connecting in a sealable relationship said upper rim of said first lower pressure vessel jacket to the outer surface of said first pressure vessel containment assembly, a second upper pressure vessel jacket adapted to enclose said first upper pressure vessel jacket and be spaced apart therefrom, said second upper pressure vessel jacket having an upper rim and a lower rim, each of said rims adapted to slidably engage the outer surface of said first upper pressure vessel jacket, means for sealing said rims, a second lower pressure vessel jacket adapted to enclose said first lower pressure vessel jacket and be spaced apart therefrom

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)

An Approach for Selection of Flow Regime and Models for Conservative Evaluation of a Vessel Integrity Monitoring System for Water-Cooled Vacuum Vessels

International Nuclear Information System (INIS)

Pointer, W. David; Ruggles, Arthur E.

2003-01-01

Thin-walled vacuum containment vessels cooled by circulating water jackets are often utilized in research and industrial applications where isolation of equipment or experiments from the influences of the surrounding environment is desirable. The development of leaks in these vessels can result in costly downtime for the facility. A Vessel Integrity Monitoring System (VIMS) is developed to detect leak formation and estimate the size of the leak to allow evaluation of the risk associated with continued operation. A wide range of leak configurations and fluid flow phenomena are considered in the evaluation of the rate at which a tracer gas dissolved in the cooling jacket water is transported into the vacuum vessel. A methodology is presented that uses basic fluid flow models and careful evaluation of their ranges of applicability to provide a conservative estimate of the transport rates for the tracer gas and hence the time required for the VIMS to detect a leak of a given size

Design criteria for the structural analysis of shipping cask containment vessels

International Nuclear Information System (INIS)

Anon.

1979-01-01

10 CFR Part 71, Sections 71.35 and 71.36, require that packages used to transport radioactive materials meet specified normal and hypothetical accident conditions. Acceptable design criteria are presented for use in the structural analysis of the containment vessels of Type B packages used to transport irradiated nuclear fuel. Alternative design criteria meeting the structural requirements of 10 CFR Part 71, Section 71.35 and 71.36, may also be used

Spallation impact analysis of plutonium storage container at K-Area

International Nuclear Information System (INIS)

Gong, C.

2000-01-01

A 100-pound concrete block falls 55-foot from ceiling spallation upon the top of the 9975 shipping package. This finite element analysis aims to evaluate the dynamic impact from the spallation upon the packaging. The geometric configuration of the packaging is meticulously modeled in detail. However, the drum is eliminated and the fiberboard with radius greater than 5.6 inches is conservatively omitted. The primary containment vessel and 3013 container were not included to simplify the model. The concrete block is modeled as a rigid body. The material properties are conservatively selected. The final results indicate that the secondary containment vessel is intact during this spallation impact. Consequently the primary containment vessel and 3013 container would not experience damage and containment is maintained. The secondary containment vessel protects the primary containment vessel from the dynamic impact. The top fiberboard is compressed from 3.5 inches to 0.875 inches will eventually recover to 1.8 inches according to tests performed at Savannah River Technology Center (SRTC)

Device for removing hydrogen gas from the safety containment vessel of a nuclear reactor

International Nuclear Information System (INIS)

Stiefel, M.

1983-01-01

The safe processing of all concentrations of gas mixtures should be possible with such a device using a thermal recombiner of compact construction. A recombiner consisting of a metal case and diverter sheets situated in it is heated by induction. The incoming pipe for the gas mixture enriched with hydrogen and the outgoing pipe for the gas mixture with low hydrogen content are connected together by a three way valve. The third connection to the safety valve takes the larger port of the gas mixture with low hydrogen content back to the safety containment vessel. Sufficient amount of the gas mixture with low hydrogen content is taken via the three way valve to the safety containment vessel to ensure that the hydrogen content of the gas mixture taken to the recombiner remains below the 4% by volume limit. (orig./PW)

Rupture tests with reactor pressure vessel head models

International Nuclear Information System (INIS)

Talja, H.; Keinaenen, H.; Hosio, E.; Pankakoski, P.H.; Rahka, K.

2003-01-01

In the LISSAC project (LImit Strains in Severe ACcidents), partly funded by the EC Nuclear Fission and Safety Programme within the 5th Framework programme, an extensive experimental and computational research programme is conducted to study the stress state and size dependence of ultimate failure strains. The results are aimed especially to make the assessment of severe accident cases more realistic. For the experiments in the LISSAC project a block of material of the German Biblis C reactor pressure vessel was available. As part of the project, eight reactor pressure vessel head models from this material (22 NiMoCr 3 7) were tested up to rupture at VTT. The specimens were provided by Forschungszentrum Karlsruhe (FzK). These tests were performed under quasistatic pressure load at room temperature. Two specimens sizes were tested and in half of the tests the specimens contain holes describing the control rod penetrations of an actual reactor pressure vessel head. These specimens were equipped with an aluminium liner. All six tests with the smaller specimen size were conducted successfully. In the test with the large specimen with holes, the behaviour of the aluminium liner material proved to differ from those of the smaller ones. As a consequence the experiment ended at the failure of the liner. The specimen without holes yielded results that were in very good agreement with those from the small specimens. (author)

Capacity assessment of concrete containment vessels subjected to aircraft impact

Energy Technology Data Exchange (ETDEWEB)

Andonov, Anton, E-mail: anton.andonov@mottmac.com; Kostov, Marin; Iliev, Alexander

2015-12-15

Highlights: • An approach to assess the containment capacity to aircraft impact via fragility curves is proposed. • Momentum over Area was defined as most suitable reference parameter to describe the aircraft load. • The effect of the impact induced damages on the containment pressure capacity has been studied. • The studied containment shows no reduction of the pressure capacity for the investigated scenarios. • The effectiveness of innovative protective structure against aircraft impact has been evaluated. - Abstract: The paper describes the procedure and the results from the assessment of the vulnerability of a generic pre-stressed containment structure subjected to a large commercial aircraft impact. Impacts of Boeing 737, Boeing 767 and Boeing 747 have been considered. The containment vulnerability is expressed by fragility curves based on the results of a number of nonlinear dynamic analyses. Three reference parameters have been considered as impact intensity measure in the fragility curve definition: peak impact force (PIF), peak impact pressure (PIP) and Momentum over Area (MoA). Conclusions on the most suitable reference parameter as well on the vulnerability of such containment vessels are drawn. The influence of the aircraft impact induced damages on the containment ultimate pressure capacity is also assessed and some preliminary conclusions on this are drawn. The paper also addresses a conceptual design of a protective structure able to decrease the containment vulnerability and provide a preliminary assessment of the applicability of such concept.

Capacity assessment of concrete containment vessels subjected to aircraft impact

International Nuclear Information System (INIS)

Andonov, Anton; Kostov, Marin; Iliev, Alexander

2015-01-01

Highlights: • An approach to assess the containment capacity to aircraft impact via fragility curves is proposed. • Momentum over Area was defined as most suitable reference parameter to describe the aircraft load. • The effect of the impact induced damages on the containment pressure capacity has been studied. • The studied containment shows no reduction of the pressure capacity for the investigated scenarios. • The effectiveness of innovative protective structure against aircraft impact has been evaluated. - Abstract: The paper describes the procedure and the results from the assessment of the vulnerability of a generic pre-stressed containment structure subjected to a large commercial aircraft impact. Impacts of Boeing 737, Boeing 767 and Boeing 747 have been considered. The containment vulnerability is expressed by fragility curves based on the results of a number of nonlinear dynamic analyses. Three reference parameters have been considered as impact intensity measure in the fragility curve definition: peak impact force (PIF), peak impact pressure (PIP) and Momentum over Area (MoA). Conclusions on the most suitable reference parameter as well on the vulnerability of such containment vessels are drawn. The influence of the aircraft impact induced damages on the containment ultimate pressure capacity is also assessed and some preliminary conclusions on this are drawn. The paper also addresses a conceptual design of a protective structure able to decrease the containment vulnerability and provide a preliminary assessment of the applicability of such concept.

An Experimental Study Of The Stability Of Vessel-Spanning Bubbles In Cylindrical, Annular, Obround and Conical Containers

International Nuclear Information System (INIS)

Dhaliwal, T.K.

2010-01-01

This report provides a summary of experiments that were performed by Fauske and Associates on the stability of vessel-spanning bubbles. The report by Fauske and Associates, An Experimental Study of the Stability of Vessel-Spanning Bubbles in Cylindrical, Annular, Obround and Conical Containers, is included in Appendix A. Results from the experiments confirm that the gravity yield parameter, Y G , correctly includes container size and can be applied to full-scale containers to predict the possibility of the formation of a stable vessel spanning bubble. The results also indicate that a vessel spanning bubble will likely form inside the STSC for KE, KW, and Settler sludges if the shear strengths of these sludges exceed 1820, 2080, and 2120 Pa, respectively. A passive mechanism installed in the STSC is effective at disrupting a rising sludge plug and preventing the sludge from plugging the vent filter or being forced out of the container. The Sludge Treatment Project for Engineered Container and Settler Sludge (EC/ST) Disposition Subproject is being conducted in two phases. Phase 1 of the EC/ST Disposition Subproject will retrieve the radioactive sludge currently stored in the K West (KW) Basin into Sludge Transport and Storage Containers (STSCs) and transport the STSCs to T-Plant for interim storage. Phase 2 of the EC/ST Disposition Subproject will retrieve the sludge from interim storage, treat and package sludge for disposal at the Waste Isolation Pilot Plant. The STSC is a cylindrical container; similar to previously used large diameter containers. A STSC (Figure 1) with a diameter of 58 inches will be used to transport KE and KW originating sludge (located in Engineered Containers 210, 220, 240, 250, and 260) to T-Plant. A STSC with an annulus (Figure 2) will be used to transport Settler Tank sludge, located in Engineered Container 230. An obround small canister design was previously considered to retrieve sludge from the basin. The obround design was selected in

A stochastic-bayesian model for the fracture probability of PWR pressure vessels

Energy Technology Data Exchange (ETDEWEB)

Francisco, Alexandre S.; Duran, Jorge Alberto R., E-mail: afrancisco@metal.eeimvr.uff.br, E-mail: duran@metal.eeimvr.uff.br [Universidade Federal Fluminense (UFF), Volta Redonda, RJ (Brazil). Dept. de Engenharia Mecanica

2013-07-01

Fracture probability of pressure vessels containing cracks can be obtained by methodologies of easy understanding, which require a deterministic treatment, complemented by statistical methods. However, more accurate results are required, methodologies need to be better formulated. This paper presents a new methodology to address this problem. First, a more rigorous methodology is obtained by means of the relationship of probability distributions that model crack incidence and nondestructive inspection efficiency using the Bayes' theorem. The result is an updated crack incidence distribution. Further, the accuracy of the methodology is improved by using a stochastic model for the crack growth. The stochastic model incorporates the statistical variability of the crack growth process, combining the stochastic theory with experimental data. Stochastic differential equations are derived by the randomization of empirical equations. From the solution of this equation, a distribution function related to the crack growth is derived. The fracture probability using both probability distribution functions is in agreement with theory, and presents realistic value for pressure vessels. (author)

A stochastic-bayesian model for the fracture probability of PWR pressure vessels

International Nuclear Information System (INIS)

Francisco, Alexandre S.; Duran, Jorge Alberto R.

2013-01-01

Fracture probability of pressure vessels containing cracks can be obtained by methodologies of easy understanding, which require a deterministic treatment, complemented by statistical methods. However, more accurate results are required, methodologies need to be better formulated. This paper presents a new methodology to address this problem. First, a more rigorous methodology is obtained by means of the relationship of probability distributions that model crack incidence and nondestructive inspection efficiency using the Bayes' theorem. The result is an updated crack incidence distribution. Further, the accuracy of the methodology is improved by using a stochastic model for the crack growth. The stochastic model incorporates the statistical variability of the crack growth process, combining the stochastic theory with experimental data. Stochastic differential equations are derived by the randomization of empirical equations. From the solution of this equation, a distribution function related to the crack growth is derived. The fracture probability using both probability distribution functions is in agreement with theory, and presents realistic value for pressure vessels. (author)

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

Containment vessel bottom head transport and lifting technique

International Nuclear Information System (INIS)

Zheng Donghong; Tian Shiyong; Hu Dequan; Xiao Hongtao

2013-01-01

The challengeable transport and lifting techniques and high safety assurance measures are needed for the onsite construction of the AP1000 containment vessel bottom head (CVBH), which is a large component with heavy weight, big size, high center of gravity, and easy to deformation. During transport, the infra structural road foundation is heavily loaded with big turning radius, and the requirement for synchronization of transport vehicles is strict. During lifting, the crane lifting capacities are high, requirement for the lifting and rigging tools is strict, nuclear island being put into place is difficult, and the crane operating foundation is heavily loaded. The transport and lifting techniques and safety assurance measures for CVBH are elaborated in detail, so as to provide a reference for the follow-up transport and lifting of large components of nuclear island. (authors)

Two, three and four buttressed PWR containment vessels. A comparative study

International Nuclear Information System (INIS)

Dufour, C.J.; Cheyrezy, M.H.; Thorsen, N.E.

1977-01-01

The purpose of the paper is to analyse the advantages and drawbacks of different arrangements of hood tendons and buttresses for a PWR containment vessel. It is shown that the solution with two buttresses and full hoop tendons (through 360 0 ) gives: acceptable secondary stresses and strains; a lower total cost of the prestressing; a shorter time schedule for prestressing operations; a smaller 'forbidden area' for the penetration lay out. This arrangement has been used for the first time (to our knowledge) in PWR containment design for the nuclear plant of DOEL III, Belgium. The comparative study has been carried out for a containment vessel with a 85 cm thick wall and an interior diameter of 42.50 m. The Guaranteed Ultimate Tensile Strength of the tendons is 9000 KN. The minimum required prestressing force is 10 300 KN per linear meter of height. The buttresses are 4.0 m long and 1.70 m thick on their vertical centerline. The following arrangements have been studied: 4 buttresses and 3/4 hoop tendons; 3 buttresses and 2/3 hoop tendons; 3 buttresses and full hoop tendons; 2 buttresses and full hoop tendons. The ovalisation and other secondary effects result mainly from three different causes: the average prestressing force is not constant around a hoop; the buttress itself constitutes a sudden thickening and creates local disturbances even under an axisymmetric loading: the anchored tendons are straight over their first 6 meters and they initiate local stress perturbations in the vicinity of the junction of the buttress and the wall. The determination of these secondary effects has been performed by a plane stress finite element analysis

Hierarchical and coupling model of factors influencing vessel traffic flow.

Science.gov (United States)

Liu, Zhao; Liu, Jingxian; Li, Huanhuan; Li, Zongzhi; Tan, Zhirong; Liu, Ryan Wen; Liu, Yi

2017-01-01

Understanding the characteristics of vessel traffic flow is crucial in maintaining navigation safety, efficiency, and overall waterway transportation management. Factors influencing vessel traffic flow possess diverse features such as hierarchy, uncertainty, nonlinearity, complexity, and interdependency. To reveal the impact mechanism of the factors influencing vessel traffic flow, a hierarchical model and a coupling model are proposed in this study based on the interpretative structural modeling method. The hierarchical model explains the hierarchies and relationships of the factors using a graph. The coupling model provides a quantitative method that explores interaction effects of factors using a coupling coefficient. The coupling coefficient is obtained by determining the quantitative indicators of the factors and their weights. Thereafter, the data obtained from Port of Tianjin is used to verify the proposed coupling model. The results show that the hierarchical model of the factors influencing vessel traffic flow can explain the level, structure, and interaction effect of the factors; the coupling model is efficient in analyzing factors influencing traffic volumes. The proposed method can be used for analyzing increases in vessel traffic flow in waterway transportation system.

Study of ex-vessel steam explosion risk of Reactor Pit Flooding System and structural response of containment for CPR1000"+ Unit

International Nuclear Information System (INIS)

Zhang Juanhua; Chen Peng

2015-01-01

Reactor Pit Flooding System is one of the special mitigation measures for severe accident for CPR1000"+ Unit. If the In-Vessel Relocation function of Reactor Pit Flooding System is failed, there is the steam explosion risk in reactor cavity. This paper firstly adopts MC3D code to build steam explosion model in order to calculate the pressure load and impulses of steam explosion that are as the input data of containment structural response analysis. The next step is to model the containment structure and analyze the structural response by ABAQUS code. The analysis results show that the integral damage induced by steam explosion to the external containment wall is shallow, and the containment structural integrity can be maintained. The risk and damage to the containment integrity reduced by steam explosion of RPF is small, and it does not influence the design and implementation of RPF. (author)

Quantitative determination of impurities contained in the pressure vessel of the Garigliano reactor

International Nuclear Information System (INIS)

Peselli, M.

1984-01-01

For dismantling the vessel of the Garigliano power plant, an element of fundamental importance is the evaluation of the γ activity induced in structural materials by neutron activation. With this knowledge, the most adequate cutting techniques, protection system, transport devices and final disposal can be chosen, in order to reduce the risk to both workers and population, taking into account the economical point of views. In this report, the model used for the activity estimation and the obtained results, showing good agreement with some experimental data, are described. The task was performed in the following steps: - measurements of the vessel steel composition, - evaluation of neutron flux affecting the vessel, - evaluation of the activity due to neutron flux, - data inventory from activity measurement performed on in core irradiated vessel specimens, - comparison between measurement and calculation data

A three-dimensional rupture analysis of steel liners anchored to concrete pressure and containment vessels

International Nuclear Information System (INIS)

Bangash, Y.

1987-01-01

Steel liners or plates are anchored to concrete pressure and containment vessels for nuclear and offshore facilities. Due to extreme loading conditions a liner may buckle due to the pull-out or shearing of anchors from the base metal and concrete. Under certain conditions attributed to loadings, liner metal deterioration and cracking of concrete behind the liner, the liner may fail by rupture. This paper presents a three-dimensional analysis of steel-concrete elements, using finite elements analysis in which a provision is made for liner instability, anchor strength and stiffness, concrete cracking and finally liner rupture. The analysis is tested first on an octagonal slab with and without an anchored steel liner. It is then extended to concrete pressure and containment vessels. The analytical results obtained are compared well with those available from the experimental tests and other sources. (author)

Measured Prestress Loss of over 20-Year-Old Prestressed Concrete Containment Vessels

Energy Technology Data Exchange (ETDEWEB)

Hahm, Dae Gi; Choun, Young Sun; Choi, In Kil [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2010-10-15

Most nuclear reactors, both in Korea and worldwide, are enclosed by a prestressed concrete containment vessels(PCCVs). The containment wall is approximately 1 m thick and is prestressed in two directions by large prestressing tendons. The main purpose of the containment is to maintain the structural integrity of the containment in the event of a major internal accident. The main accidental scenario, which the containment is designed to withstand, is a so-called loss of coolant accident (LOCA). A LOCA is initiated by a pipe rupture in the cooling system, discharging hot steam into the containment. The escape of steam increases both the temperature and pressure inside the containment. The increased internal pressure arising from a LOCA is referred to as the design pressure. The prestressing system is designed to counterbalance the tensile forces arising from the design pressure. The status of the containment is gradually changed due to environmental factors and by alterations in the micro structure of the material. The prestress will be reduced due to shrinkage and creep in the concrete and relaxation in the tendons. The corrosion protection of tendons are for Korean containments arranged in two different ways, either by cement grouting (bonded tendons) or e.g. by grease injection (unbonded tendons). The major advantage using unbonded tendons is the possibilities of assessing their status (e.g. prestress losses or corrosion damages) which is not possible using bonded tendons. Both bonded and unbonded tendons are used worldwide. For example in the U.S. almost all tendons are unbonded, whereas in France almost all tendons are bonded. For Korean reactor containments with unbonded tendons (14 containments) the tendon force is monitored at regular in-service inspections. The power plant Wolsung in Korea has bonded tendons and several prestressed concrete beams were constructed with the single purpose to follow up the prestress losses. The remaining tendon forces in some

Measured Prestress Loss of over 20-Year-Old Prestressed Concrete Containment Vessels

International Nuclear Information System (INIS)

Hahm, Dae Gi; Choun, Young Sun; Choi, In Kil

2010-01-01

Most nuclear reactors, both in Korea and worldwide, are enclosed by a prestressed concrete containment vessels(PCCVs). The containment wall is approximately 1 m thick and is prestressed in two directions by large prestressing tendons. The main purpose of the containment is to maintain the structural integrity of the containment in the event of a major internal accident. The main accidental scenario, which the containment is designed to withstand, is a so-called loss of coolant accident (LOCA). A LOCA is initiated by a pipe rupture in the cooling system, discharging hot steam into the containment. The escape of steam increases both the temperature and pressure inside the containment. The increased internal pressure arising from a LOCA is referred to as the design pressure. The prestressing system is designed to counterbalance the tensile forces arising from the design pressure. The status of the containment is gradually changed due to environmental factors and by alterations in the micro structure of the material. The prestress will be reduced due to shrinkage and creep in the concrete and relaxation in the tendons. The corrosion protection of tendons are for Korean containments arranged in two different ways, either by cement grouting (bonded tendons) or e.g. by grease injection (unbonded tendons). The major advantage using unbonded tendons is the possibilities of assessing their status (e.g. prestress losses or corrosion damages) which is not possible using bonded tendons. Both bonded and unbonded tendons are used worldwide. For example in the U.S. almost all tendons are unbonded, whereas in France almost all tendons are bonded. For Korean reactor containments with unbonded tendons (14 containments) the tendon force is monitored at regular in-service inspections. The power plant Wolsung in Korea has bonded tendons and several prestressed concrete beams were constructed with the single purpose to follow up the prestress losses. The remaining tendon forces in some

Hierarchical and coupling model of factors influencing vessel traffic flow.

Directory of Open Access Journals (Sweden)

Zhao Liu

Full Text Available Understanding the characteristics of vessel traffic flow is crucial in maintaining navigation safety, efficiency, and overall waterway transportation management. Factors influencing vessel traffic flow possess diverse features such as hierarchy, uncertainty, nonlinearity, complexity, and interdependency. To reveal the impact mechanism of the factors influencing vessel traffic flow, a hierarchical model and a coupling model are proposed in this study based on the interpretative structural modeling method. The hierarchical model explains the hierarchies and relationships of the factors using a graph. The coupling model provides a quantitative method that explores interaction effects of factors using a coupling coefficient. The coupling coefficient is obtained by determining the quantitative indicators of the factors and their weights. Thereafter, the data obtained from Port of Tianjin is used to verify the proposed coupling model. The results show that the hierarchical model of the factors influencing vessel traffic flow can explain the level, structure, and interaction effect of the factors; the coupling model is efficient in analyzing factors influencing traffic volumes. The proposed method can be used for analyzing increases in vessel traffic flow in waterway transportation system.

A model for ultrasound contrast agent in a phantom vessel

KAUST Repository

Qamar, Adnan

2014-02-01

A theoretical framework to model the dynamics of Ultrasound Contrast Agent (UCA) inside a phantom vessel is presented. The model is derived from the reduced Navier-Stokes equation and is coupled with the evolving flow field solution inside the vessel by a similarity transformation approach. The results are computed, and compared with experiments available in literature, for the initial UCA radius of Ro=1.5 μm and 2 μm for the vessel diameter of D=12 μm and 200 μm with the acoustic parameters as utilized in the experiments. When compared to other models, better agreement on smaller vessel diameter is obtained with the proposed coupled model. The model also predicts, quite accurately, bubble fragmentation in terms of acoustic and geometric parameters. © 2014 IEEE.

A probability model for the failure of pressure containing parts

International Nuclear Information System (INIS)

Thomas, H.M.

1978-01-01

The model provides a method of estimating the order of magnitude of the leakage failure probability of pressure containing parts. It is a fatigue based model which makes use of the statistics available for both specimens and vessels. Some novel concepts are introduced but essentially the model simply quantifies the obvious i.e. that failure probability increases with increases in stress levels, number of cycles, volume of material and volume of weld metal. A further model based on fracture mechanics estimates the catastrophic fraction of leakage failures. (author)

Creep deformation and crack growth in a low alloy steel welded pressure vessel containing defects

International Nuclear Information System (INIS)

Coleman, M.C.

1982-01-01

A full-size pressure vessel was tested for effects of welding residual stresses on creep deformation and crack growth. The vessel, based on 1/2 Cr 1/2 Mo 1/4 V main steam pipe, contained four 2CrMo manual metal arc welds, two in the as-welded condition and two stress-relieved. All the welds contained pre-existing defects machined in the heat affected zones. Testing was carried out at two internal steam pressures, 250 and 350 bar, and 565 0 C. Cracked and uncracked areas of the vessel were monitored continuously. Results are presented for the continuous creep deformation observed in both the hoop and axial directions of the welds throughout the 11,400 h of testing, as well as the intermittent strain data obtained during inspections. Crack growth observations are described based on nondestructive examination. The residual stresses measured are also given for both the as-welded and stress relieved weldments. Results obtained are discussed in terms of the effects of welding residual stress on the hoop and axial deformations observed in the welds. Similarly, the effects of residual stress on creep crack growth are considered together with compositional and microstructural implications. 9 figures, 5 tables

Numerical modeling of the waves evolution generated by the depressurization of the vessels containing a supercritical parameters coolant

Science.gov (United States)

Alekseev, Maksim V.; Vozhakov, Ivan S.; Lezhnin, Sergey I.; Pribaturin, Nikolay A.

2017-10-01

The development of power plants focuses on increasing the parameters of water coolants up to a supercritical level. Depressurization of the unit circuits with such a coolant leads to emergency situations. Their scenarios can change significantly with the variation of initial pressure and temperature before the start of depressurization. When the pressure drops from the supercritical single-phase region of the initial thermodynamic parameters of the coolant, either the liquid boils up, or the vapor is condensed. Because of the rapid pressure decrease, the phase transition can be non-equilibrium that must be taken into account in the simulation. In the present study, an axisymmetric problem of the outflow of a water coolant from the pipe butt-end is considered. The equations of continuity, momentum and energy for a two-phase homogeneous mixture are solved numerically. The vapor and liquid properties are calculated using the TTSE software package (The Tabular Taylor Series Expansion Method). On the basis of the computer complex LCPFCT (The Flux-Corrected Transport Algorithm) the program code was developed for solving numerous problems on the depressurization of vessels or pipelines, containing superheated water or gas under high pressure. Different variants of outflow in the external model atmosphere and generation of waves are analyzed. The calculated data on the interaction of pressure waves with a barrier are calculated. To describe phase transitions, an asymptotic relaxation model of nonequilibrium evaporation and condensation has been created and tested.

Eulerian finite-difference calculations of explosions in partially water-filled overstrong cylindrical containment vessels

International Nuclear Information System (INIS)

Thompson, S.L.; Herrmann, W.

1977-01-01

Calculations, using the two-dimensional Eulerian finite-difference code CSQ, were performed for the problem of a small spherical high-explosive charge detonated in a closed heavy-walled cylindrical container partially filled with water. Data from corresponding experiments, specifically performed to validate codes used for hypothetical core disruptive accidents of liquid metal fast breeder reactors, are available in the literature. The calculations were performed specifically to test whether Eulerian methods could handle this type of problem, to determine whether water cavitation, which plays a large role in the loadings on the roof of the containment vessel, could be described adequately by an equilibrium liquid-vapor mixed phase model, and to investigate the trade-off between accuracy and cost of the calculations by using different sizes of computational meshes. Comparison of the experimental and computational data shows that the Eulerian method can handle the problem with ease, giving good predictions of wall and floor loadings. While roof loadings are qualitatively correct, peak impulse appears to be affected by numerical resolution and is underestimated somewhat

The influence of selected containment structures on debris dispersal and transport following high pressure melt ejection from the reactor vessel

International Nuclear Information System (INIS)

Pilch, M.; Tarbell, W.W.; Brockmann, J.E.

1988-09-01

High pressure expulsion of molten core debris from the reactor pressure vessel may result in dispersal of the debris from the reactor cavity. In most plants, the cavity exits into the containment such that the debris impinges on structures. Retention of the debris on the structures may affect the further transport of the debris throughout the containment. Two tests were done with scaled structural shapes placed at the exit of 1:10 linear scale models of the Zion cavity. The results show that the debris does not adhere significantly to structures. The lack of retention is attributed to splashing from the surface and reentrainment in the gas flowing over the surface. These processes are shown to be applicable to reactor scale. A third experiment was done to simulate the annular gap between the reactor vessel and cavity wall. Debris collection showed that the fraction of debris exiting through the gap was greater than the gap-to-total flow area ratio. Film records indicate that dispersal was primarily by entrainment of the molten debris in the cavity. 29 refs., 36 figs., 11 tabs

Residual stress concerns in containment analysis

International Nuclear Information System (INIS)

Costantini, F.; Kulak, R. F.; Pfeiffer, P. A.

1997-01-01

The manufacturing of steel containment vessels starts with the forming of flat plates into curved plates. A steel containment structure is made by welding individual plates together to form the sections that make up the complex shaped vessels. The metal forming and welding process leaves residual stresses in the vessel walls. Generally, the effect of metal forming residual stresses can be reduced or virtually eliminated by thermally stress relieving the vesseL In large containment vessels this may not be practical and thus the residual stresses due to manufacturing may become important. The residual stresses could possibly tiect the response of the vessel to internal pressurization. When the level of residual stresses is significant it will affect the vessel's response, for instance the yielding pressure and possibly the failure pressure. The paper will address the effect of metal forming residual stresses on the response of a generic pressure vessel to internal pressurization. A scoping analysis investigated the effect of residual forming stresses on the response of an internally pressurized vessel. A simple model was developed to gain understanding of the mechanics of the problem. Residual stresses due to the welding process were not considered in this investigation

Microstructural characterization of atom clusters in irradiated pressure vessel steels and model alloys

International Nuclear Information System (INIS)

Auger, P.; Pareige, P.; Akamatsu, M.; Van Duysen, J.C.

1993-01-01

In order to characterize the microstructural evolution of iron solid solution under irradiation, two pressure vessel steels irradiated in service conditions, and, for comparison, low copper model alloys irradiated with neutrons and electrons, have been studied through small angle neutron scattering and atom probe experiments. In Fe-Cu model alloys, copper clusters are formed containing uncertain proportions of iron. In the low copper industrial steels, the feature is more complex; solute atoms such as Ni, Mn and Si, sometimes associated with Cu, segregate as ''clouds'' more or less condensed in the iron solid solution. These silicides, or at least Si, Ni, Mn association, may facilitate the copper segregation although the initial iron matrix contains a low copper concentration. (authors). 24 refs., 3 figs., 2 tabs

Microstructural characterization of atom clusters in irradiated pressure vessel steels and model alloys

Energy Technology Data Exchange (ETDEWEB)

Auger, P; Pareige, P [Rouen Univ., 76 - Mont-Saint-Aignan (France); Akamatsu, M; Van Duysen, J C [Electricite de France (EDF), 77 - Ecuelles (France)

1994-12-31

In order to characterize the microstructural evolution of iron solid solution under irradiation, two pressure vessel steels irradiated in service conditions, and, for comparison, low copper model alloys irradiated with neutrons and electrons, have been studied through small angle neutron scattering and atom probe experiments. In Fe-Cu model alloys, copper clusters are formed containing uncertain proportions of iron. In the low copper industrial steels, the feature is more complex; solute atoms such as Ni, Mn and Si, sometimes associated with Cu, segregate as ``clouds`` more or less condensed in the iron solid solution. These silicides, or at least Si, Ni, Mn association, may facilitate the copper segregation although the initial iron matrix contains a low copper concentration. (authors). 24 refs., 3 figs., 2 tabs.

An international survey of in-service inspection experience with prestressed concrete pressure vessels and containments for nuclear reactors

International Nuclear Information System (INIS)

1982-04-01

An international survey is presented of experience obtained from the in-service surveillance of prestressed concrete pressure vessels and containments for nuclear reactors. Some information on other prestressed concrete structures is also given. Experience has been gained during the working life of such structures in Western Europe and the USA over the years since 1967. For each country a summary is given of the nuclear programme, national standards and Codes of Practice, and the detailed in-service inspection programme. Reports are then given of the actual experience obtained from the inspection programme and the methods of measurement, examination and reporting employed in each country. A comprehensive bibliography of over 100 references is included. The appendices contain information on nuclear power stations which are operating, under construction or planned worldwide and which employ either prestressed concrete pressure vessels or containments. (U.K.)

Report of Task Group on Ex-Vessel Thermal-Hydraulics Corium/concrete interactions and combustible gas distribution in large dry containments

International Nuclear Information System (INIS)

1987-11-01

The Task Group on Ex-Vessel Thermal-Hydraulics was established by the PWG 2 to address the physical processes that occur in the ex-vessel phase of severe accidents, to study their impact on containment loading and failure, and to assess the available calculation methods. This effort is part of an overall CSNI effort to come to an international understanding of the issues involved. The Task Group decided to focus its initial efforts on the Large Dry Containment used extensively to contain the consequences of postulated (design basis) accidents in Light Water Reactors (LWR). Although such containments have not been designed with explicit consideration of severe accidents, recent assessments indicate a substantial inherent capability for these accidents. The Task Group has examined the loads likely to challenge the integrity of the containment, and considered the calculation of the containment's response. This report is the outcome of this effort

Vessel Operating Units (Vessels)

Data.gov (United States)

National Oceanic and Atmospheric Administration, Department of Commerce — This data set contains data for vessels that are greater than five net tons and have a current US Coast Guard documentation number. Beginning in1979, the NMFS...

Marine Vessel Models in Changing Operational Conditions - A Tutorial

DEFF Research Database (Denmark)

Perez, Tristan; Sørensen, Asgeir; Blanke, Mogens

2006-01-01

conditions (VOC). However, since marine systems operate in changing VOCs, there is a need to adapt the models. To date, there is no theory available to describe a general model valid across different VOCs due to the complexity of the hydrodynamic involved. It is believed that system identification could......This tutorial paper provides an introduction, from a systems perspective, to the topic of ship motion dynamics of surface ships. It presents a classification of parametric models currently used for monitoring and control of marine vessels. These models are valid for certain vessel operational...

Stress analysis of R2 pressure vessel. Structural reliability benchmark exercise

International Nuclear Information System (INIS)

Vestergaard, N.

1987-05-01

The Structural Reliability Benchmark Exercise (SRBE) is sponsored by the EEC as part of the Reactor Safety Programme. The objectives of the SRBE are to evaluate and improve 1) inspection procedures, which use non-destructive methods to locate defects in pressure (reactor) vessels, as well as 2) analytical damage accumulation models, which predict the time to failure of vessels containing defects. In order to focus attention, an experimental presure vessel has been inspected, subjected fatigue loadings and subsequently analysed by several teams using methods of their choice. The present report contains the first part of the analytical damage accumulation analysis. The stress distributions in the welds of the experimental pressure vessel were determined. These stress distributions will be used to determine the driving forces of the damage accumulation models, which will be addressed in a future report. (author)

Accuracy of geometrical modelling of heat transfer from tissue to blood vessels

NARCIS (Netherlands)

Leeuwen, van G.M.J.; Kotte, A.N.T.J.; Bree, de J.; Koijk, van der J.F.; Crezee, J.; Lagendijk, J.J.W.

1997-01-01

We have developed a thermal model in which blood vessels are described as geometrical objects, 3D curves with associated diameters. Here the behaviour of the model is examined for low resolutions compared with the vessel diameter and for strongly curved vessels. The tests include a single straight

Development of a method for detecting nuclear fuel debris and water leaks at a nuclear reactor/containment vessel by flow visualization

International Nuclear Information System (INIS)

Umezawa, Shuichi; Tanaka, Katsuhiko

2013-01-01

It is the important issue to fill up each nuclear reactor/containment vessel with water and to take out debris of damaged fuel from them for decommissioning of Fukushima Daiichi nuclear power plants. It is necessary to detect the debris and water leaks at a nuclear reactor/containment vessel for the purpose. However, the method is not completely developed in the present stage. Accordingly, we have developed a method for detecting debris and water leaks at a nuclear reactor/containment vessel by flow visualization. Experiments of the flow visualization were conducted using two types of water tanks. An optical fiber and a collimator lens were employed for modifying a straight laser beam into a sheet projection. Some visualized images were obtained through the experiments. Particle Image Velocimetry, i.e. PIV, analysis was applied to the images for quantitative flow rate analysis. Consequently, it is considered that the flow visualization method has a possibility for the practical use. (author)

Emergency venting of pressure vessels

International Nuclear Information System (INIS)

Steinkamp, H.

1995-01-01

With the numerical codes developed for safety analysis the venting of steam vessel can be simulated. ATHLET especially is able to predict the void fraction depending on the vessel height. Although these codes contain a one-dimensional model they allow the description of complex geometries due to the detailed nodalization of the considered apparatus. In chemical reactors, however, the venting process is not only influenced by the flashing behaviour but additionally by the running chemical reaction in the vessel. Therefore the codes used for modelling have to consider the kinetics of the chemical reaction. Further multi-component systems and dissolving processes have to be regarded. In order to preduct the fluid- and thermodynamic process it could be helpful to use 3-dimensional codes in combination with the one-dimensional codes as used in nuclear industry to get a more detailed describtion of the running processes. (orig./HP)

Test results on direct containment heating by high-pressure melt ejection into the Surtsey vessel: The TDS test series

International Nuclear Information System (INIS)

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

1994-08-01

The Technology Development and Scoping (TDS) test series was conducted to test and develop instrumentation and procedures for performing steam-driven, high-pressure melt ejection (HPME) experiments at the Surtsey Test Facility to investigate direct containment heating (DCH). Seven experiments, designated TDS-1 through TDS-7, were performed in this test series. These experiments were conducted using similar initial conditions; the primary variable was the initial pressure in the Surtsey vessel. All experiments in this test series were performed with a steam driving gas pressure of ≅ 4 MPa, 80 kg of lumina/iron/chromium thermite melt simulant, an initial hole diameter of 4.8 cm (which ablated to a final hole diameter of ≅ 6 cm), and a 1/10th linear scale model of the Surry reactor cavity. The Surtsey vessel was purged with argon ( 2 ) to limit the recombination of hydrogen and oxygen, and gas grab samples were taken to measure the amount of hydrogen produced

Predictive simulation of bidirectional Glenn shunt using a hybrid blood vessel model.

Science.gov (United States)

Li, Hao; Leow, Wee Kheng; Chiu, Ing-Sh

2009-01-01

This paper proposes a method for performing predictive simulation of cardiac surgery. It applies a hybrid approach to model the deformation of blood vessels. The hybrid blood vessel model consists of a reference Cosserat rod and a surface mesh. The reference Cosserat rod models the blood vessel's global bending, stretching, twisting and shearing in a physically correct manner, and the surface mesh models the surface details of the blood vessel. In this way, the deformation of blood vessels can be computed efficiently and accurately. Our predictive simulation system can produce complex surgical results given a small amount of user inputs. It allows the surgeon to easily explore various surgical options and evaluate them. Tests of the system using bidirectional Glenn shunt (BDG) as an application example show that the results produc by the system are similar to real surgical results.

Significance of fluid-structure interaction phenomena for containment response to ex-vessel steam explosions

Energy Technology Data Exchange (ETDEWEB)

Almstroem, H.; Sundel, T. [National Defence Research Establishment, Stockholm (Sweden); Frid, W.; Engelbrektson, A.

1998-01-01

When studying the structural response of a containment building to ex-vessel steam explosion loads, a two-step procedure is often used. In the first step of this procedure the structures are treated as rigid and the pressure-time history generated by the explosion at the rigid wall is calculated. In the second step the calculated pressure is applied to the structures. The obvious weakness of the two-step procedure is that it does not correspond to the real dynamic behaviour of the fluid-structure system. The purpose of this paper is to identify and evaluate the relevant fluid-structure interaction phenomena. This is achieved through direct treatment of the explosion process and the structural response. The predictions of a direct and two-step treatment are compared for a BWR Mark II containment design, consisting of two concentric walls interacting with water masses in the central and annular pools. It is shown that the two-step approach leads to unrealistic energy transfer in the containment system studied, and to significant overestimation of the deflection of the containment wall. As regards the pedestal wall, the direct method analysis shows that the flexibility of this wall affects the pressure-time history considerably. Three load types have been identified for this wall namely shock load, water blow as a result of water cavitation, and hydrodynamic load. Reloading impulse due to cavitation phenomena plays an important role as it amounts to about 40% of the total impulse load. Investigation of the generality of the cavitation phenomena in the context of ex-vessel steam explosion loads was outside the scope of this work. (author)

Significance of fluid-structure interaction phenomena for containment response to ex-vessel steam explosions

Energy Technology Data Exchange (ETDEWEB)

Almstroem, H.; Sundel, T. (Nat. Defence Res. Establ., Tumba (Sweden)); Frid, W. (Swedish Nuclear Power Inspectorate, SE-10658, Stockholm (Sweden)); Engelbrektson, A. (VBB/SWECO, Box 34044, SE-10026, Stockholm (Sweden))

1999-05-01

When studying the structural response of a containment building to ex-vessel steam explosion loads, a two-step procedure is often used. In the first step of this procedure the structures are treated as rigid and the pressure-time history generated by the explosion, at the rigid wall, is calculated. In the second step the calculated pressure is applied to the structures. The obvious weakness of the two-step procedure is that it does not correspond to the real dynamic behaviour of the fluid-structure system. The purpose of this paper is to identify and evaluate the relevant fluid-structure interaction phenomena. This is achieved through direct treatment of the explosion process and the structural response. The predictions of a direct and two-step treatment are compared for a BWR Mark II containment design, consisting of two concentric walls interacting with water masses in the central and annular pools. It is shown that the two-step approach leads to unrealistic energy transfer in the containment system studied and to significant overestimation of the deflection of the containment wall. As regards the pedestal wall, the direct method analysis shows that the flexibility of this wall affects the pressure-time history considerably. Three load types have been identified for this wall namely shock load, water blow as a result of water cavitation, and hydrodynamic load. Reloading impulse due to cavitation phenomena plays an important role as it amounts to [approx]40% of the total impulse load. Investigation of the generality of the cavitation phenomena in the context of ex-vessel steam explosion loads was outside the scope of this work. (orig.) 5 refs.

Significance of fluid-structure interaction phenomena for containment response to ex-vessel steam explosions

International Nuclear Information System (INIS)

Almstroem, H.; Sundel, T.; Frid, W.; Engelbrektson, A.

1999-01-01

When studying the structural response of a containment building to ex-vessel steam explosion loads, a two-step procedure is often used. In the first step of this procedure the structures are treated as rigid and the pressure-time history generated by the explosion, at the rigid wall, is calculated. In the second step the calculated pressure is applied to the structures. The obvious weakness of the two-step procedure is that it does not correspond to the real dynamic behaviour of the fluid-structure system. The purpose of this paper is to identify and evaluate the relevant fluid-structure interaction phenomena. This is achieved through direct treatment of the explosion process and the structural response. The predictions of a direct and two-step treatment are compared for a BWR Mark II containment design, consisting of two concentric walls interacting with water masses in the central and annular pools. It is shown that the two-step approach leads to unrealistic energy transfer in the containment system studied and to significant overestimation of the deflection of the containment wall. As regards the pedestal wall, the direct method analysis shows that the flexibility of this wall affects the pressure-time history considerably. Three load types have been identified for this wall namely shock load, water blow as a result of water cavitation, and hydrodynamic load. Reloading impulse due to cavitation phenomena plays an important role as it amounts to ∼40% of the total impulse load. Investigation of the generality of the cavitation phenomena in the context of ex-vessel steam explosion loads was outside the scope of this work. (orig.)

Characterization of atom clusters in irradiated pressure vessel steels and model alloys

International Nuclear Information System (INIS)

Auger, P.; Pareige, P.; Akamatsu, M.; Van Duysen, J.C.

1993-12-01

In order to characterize the microstructural evolution of the iron solid solution under irradiation, two pressure vessel steels irradiated in service conditions and, for comparison, low copper model alloys irradiated with neutrons and electrons have been studied. The characterization has been carried out mainly thanks to small angle neutron scattering and atom probe experiments. Both techniques lead to the conclusion that clusters develop with irradiations. In Fe-Cu model alloys, copper clusters are formed containing uncertain proportions of iron. In the low copper industrial steels, the feature is more complex. Solute atoms like Ni, Mn and Si, sometimes associated with Cu, segregate as ''clouds'' more or less condensed in the iron solid solution. These silicides, or at least Si, Ni, Mn association, may facilitate the copper segregation although the initial iron matrix contains a low copper concentration. (authors). 24 refs., 3 figs., 2 tabs

Simulation of containment phenomena during the Phebus FPT1 test with the CONTAIN code

International Nuclear Information System (INIS)

Kljenak, I.; Mavko, B.

2002-01-01

Thermal-hydraulic and aerosol phenomena which occurred in the containment vessel of the Phebus integral experimental facility during the first 30000 s of the Phebus FPT1 test were simulated with the CONTAIN thermal-hydraulic computer code. A single-cell input model of the vessel was developed, and boundary and initial conditions that were determined during the experiment were applied. The comparison of experimental and calculated results shows that, although the atmosphere temperature was well simulated, the calculated condensation rate was apparently too high, resulting in a lower pressure of the containment atmosphere. The aerosol deposition process was well simulated.(author)

Evaluation of seismic shear capacity of prestressed concrete containment vessels with fiber reinforcement

Energy Technology Data Exchange (ETDEWEB)

Choun, Young Sun; Park, Jun Hee [Integrated Safety Assessment Division, Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2015-10-15

Fibers have been used in cement mixture to improve its toughness, ductility, and tensile strength, and to enhance the cracking and deformation characteristics of concrete structural members. The addition of fibers into conventional reinforced concrete can enhance the structural and functional performances of safety-related concrete structures in nuclear power plants. The effects of steel and polyamide fibers on the shear resisting capacity of a prestressed concrete containment vessel (PCCV) were investigated in this study. For a comparative evaluation between the shear performances of structural walls constructed with conventional concrete, steel fiber reinforced concrete, and polyamide fiber reinforced concrete, cyclic tests for wall specimens were conducted and hysteretic models were derived. The shear resisting capacity of a PCCV constructed with fiber reinforced concrete can be improved considerably. When steel fiber reinforced concrete contains hooked steel fibers in a volume fraction of 1.0%, the maximum lateral displacement of a PCCV can be improved by > 50%, in comparison with that of a conventional PCCV. When polyamide fiber reinforced concrete contains polyamide fibers in a volume fraction of 1.5%, the maximum lateral displacement of a PCCV can be enhanced by ∼40%. In particular, the energy dissipation capacity in a fiber reinforced PCCV can be enhanced by > 200%. The addition of fibers into conventional concrete increases the ductility and energy dissipation of wall structures significantly. Fibers can be effectively used to improve the structural performance of a PCCV subjected to strong ground motions. Steel fibers are more effective in enhancing the shear performance of a PCCV than polyamide fibers.

Evaluation of seismic shear capacity of prestressed concrete containment vessels with fiber reinforcement

International Nuclear Information System (INIS)

Choun, Young Sun; Park, Jun Hee

2015-01-01

Fibers have been used in cement mixture to improve its toughness, ductility, and tensile strength, and to enhance the cracking and deformation characteristics of concrete structural members. The addition of fibers into conventional reinforced concrete can enhance the structural and functional performances of safety-related concrete structures in nuclear power plants. The effects of steel and polyamide fibers on the shear resisting capacity of a prestressed concrete containment vessel (PCCV) were investigated in this study. For a comparative evaluation between the shear performances of structural walls constructed with conventional concrete, steel fiber reinforced concrete, and polyamide fiber reinforced concrete, cyclic tests for wall specimens were conducted and hysteretic models were derived. The shear resisting capacity of a PCCV constructed with fiber reinforced concrete can be improved considerably. When steel fiber reinforced concrete contains hooked steel fibers in a volume fraction of 1.0%, the maximum lateral displacement of a PCCV can be improved by > 50%, in comparison with that of a conventional PCCV. When polyamide fiber reinforced concrete contains polyamide fibers in a volume fraction of 1.5%, the maximum lateral displacement of a PCCV can be enhanced by ∼40%. In particular, the energy dissipation capacity in a fiber reinforced PCCV can be enhanced by > 200%. The addition of fibers into conventional concrete increases the ductility and energy dissipation of wall structures significantly. Fibers can be effectively used to improve the structural performance of a PCCV subjected to strong ground motions. Steel fibers are more effective in enhancing the shear performance of a PCCV than polyamide fibers

Nuclear reactor construction with bottom supported reactor vessel

International Nuclear Information System (INIS)

Sharbaugh, J.E.

1987-01-01

This patent describes an improved liquid metal nuclear reactor construction comprising: (a) a nuclear reactor core having a bottom platform support structure; (b) a reactor vessel for holding a large pool of low pressure liquid metal coolant and housing the core; (c) a containment structure surrounding the reactor vessel and having a sidewall spaced outwardly from the reactor vessel side wall and having a base mat spaced below the reactor vessel bottom end wall; (d) a central small diameter post anchored to the containment structure base mat and extending upwardly to the reactor vessel to axially fix the bottom end wall of the reactor vessel and provide a center column support for the lower end of the reactor core; (e) annular support structure disposed in the reactor vessel on the bottom end wall and extending about the lower end of the core; (f) structural support means disposed between the containment structure base mat and bottom end of the reactor vessel wall and cooperating for supporting the reactor vessel at its bottom end wall on the containment structure base mat to allow the reactor vessel to expand radially but substantially prevent any lateral motions that might be imposed by the occurrence of a seismic event; (g) a bed of insulating material disposed between the containment structure base mat and the bottom end wall of the reactor vessel and uniformly supporting the reactor vessel at its bottom end wall; freely expand radially from the central post as it heats up while providing continuous support thereof; (h) a deck supported upon the wall of the containment vessel above the top open end of the reactor vessel; and (i) extendible and retractable coupling means extending between the deck and the top open end of the reactor vessel and flexibly and sealably interconnecting the reactor vessel at its top end to the deck

Manufacturing method for radioactive material containing vessel

International Nuclear Information System (INIS)

Nishioka, Hideharu; Matsushita, Kazuo; Toyota, Michinori.

1997-01-01

Lead homogenization is applied on the inner surface of a space formed between an inner cylinder and an outer cylinder, and a molten lead heated to about 400 to 500degC is cast into a space formed between the inner cylinder and the outer cylinder in a state where the inner and the outer cylinders are heated to from 200 to 300degC. The space formed between the inner cylinder and the outer cylinder is heated to and kept at 330degC or higher for at least 2minutes after the casting of the molten lead, and then it is cooled. Thus, lowering of density of the molten lead due to excess elevation of temperature or dropping of the lead at the homogenization portion by heating the inner and the outer cylinders to an excessively high temperature are not caused. In addition, formation of gaps in the boundary between the inner cylinder and the outer cylinder or between the lead of the homogenized portion and that of the cast portion due to the melting of the lead of the homogenized portion in the space is prevented reliably thereby capable of forming a satisfactory shielding member. Then, even when the thickness of the inner cylinder and the outer cylinder is large, radioactive material containing vessel excellent in heat releasing property and radiation shielding property can be manufactured. (N.H.)

In-vessel retention modeling capabilities in MAAP5

International Nuclear Information System (INIS)

Paik, Chan Y.; Lee, Sung Jin; Zhou, Quan; Luangdilok, W.; Reeves, R.W.; Henry, R.E.; Plys, M.; Scobel, J.H.

2012-01-01

Modular Accident Analysis Program (MAAP) is an integrated severe accident analysis code for both light water and heavy water reactors. New and improved models to address the complex phenomena associated with in-vessel retention (IVR) were incorporated into MAAP5.01. They include: -a) time-dependent volatile and non-volatile decay heat, -b) material properties at high temperatures, -c) finer vessel wall nodalization, -d) new correlations for natural convection heat transfer in the oxidic pool, -e) refined metal layer heat transfer to the reactor vessel wall and surroundings, -f) formation of a heavy metal layer, and -g) insulation cooling channel model and associated ex-vessel heat transfer and critical heat flux correlations. In this paper, the new and improved models in MAAP5.01 are described and sample calculation results are presented for the AP1000 passive plant. For the IVR evaluation, a transient calculation is useful because the timing of corium relocation, decaying heat load, and formation of separate layers in the lower plenum all affect integrity of the lower head. The key parameters affecting the IVR success are the metal layer emissivity and thickness of the top metal layer, which depends on the amount of steel in the oxidic pool and in the heavy metal layer. With the best estimate inputs for the debris mixing parameters in a conservative IVR scenario, the AP1000 plant results show that the maximum ex-vessel heat flux to CHF ratio is about 0.7, which occurs before 10.000 seconds when the decay heat is high. The AP1000 plant results demonstrate how MAAP5.01 can be used to evaluate IVR and to gain insight into responses of the lower head during a severe accident

Endurance test report of rubber sealing materials for the containment vessel

International Nuclear Information System (INIS)

Yamamoto, R.; Watanabe, K.; Hanashima, K.

2015-01-01

In the event of a nuclear power plant accident such as a core meltdown and a cooling system failure, the containment contains radioactive materials released from the reactor pressure vessel to reduce the activity of the radioactive materials and the effects of radiation in the vicinity of the plant. Since high sealing performance and high pressure resistance are required of the containment, a silicone or EPDM rubber gasket with high heat and radiation resistance is used for the sealing of the sealing boundary of the containment. In recent years, it has been shown that a large amount of steam is released into the containment in the case of a severe accident. Consequently, radiation resistance at high temperature as well as steam resistance is required of the rubber gasket placed at the sealing boundary. However, the steam resistance of silicone rubber is not necessarily as good as that of EPDM rubber. Therefore, it is necessary to evaluate the sealing characteristics of rubber gaskets in such a degrading environment in a severe accident. O. Kato et al. [1] conducted a study on the degradation status of rubber gaskets and their application limits at high temperature. However, few studies have evaluated rubber gaskets in high-temperature radiation and steam environments. In this study, we degraded silicone rubber and EPDM rubber used for the containment in the high-temperature radiation and steam environments expected to occur in a severe accident and evaluated the useful life of the rubber as a sealing material by estimating the change in its performance as a sealing material from the change in permanent compressive strain in the rubber. (author)

Tearing stability analysis of an axial surface flaw in thick-walled pressure vessels

International Nuclear Information System (INIS)

Zahoor, A.; Ghassemi, B.B.

1991-01-01

This paper presents two fracture mechanics models for evaluation of an axial surface flaw in pressure vessels. The surface flaw is located on the outside surface of the vessel. The first model assumes yielding of the remaining ligament directly ahead of the flaw. The second model assumes contained yielding ahead of the flaw and uses a linear elastic fracture mechanics solution. The former model is suitable for cases where the combination of material toughness, flaw size, and load is such that initiation of flaw growth follows ligament yielding. The latter model is suitable for low-toughness materials where initiation of crack growth and potential tearing instability may occur prior to the yielding of the ligament. Both models are suitable for thick-walled vessels. The paper discusses the applicability regime for both models. The models are then applied to a test vessel and the predicted failure pressure is compared against the pressure attained in the test. Results show that both models can be applied successfully. In particular, the contained yielding model when used with the plane-stress assumption can give reasonable predictions even for cases that involve yielding of the ligament. (orig.)

Tearing stability analysis of an axial surface flaw in thick-walled pressure vessels

Energy Technology Data Exchange (ETDEWEB)

Zahoor, A.; Ghassemi, B.B. (NOVETECH Corp., Rockville, MD (USA))

1991-04-01

This paper presents two fracture mechanics models for evaluation of an axial surface flaw in pressure vessels. The surface flaw is located on the outside surface of the vessel. The first model assumes yielding of the remaining ligament directly ahead of the flaw. The second model assumes contained yielding ahead of the flaw and uses a linear elastic fracture mechanics solution. The former model is suitable for cases where the combination of material toughness, flaw size, and load is such that initiation of flaw growth follows ligament yielding. The latter model is suitable for low-toughness materials where initiation of crack growth and potential tearing instability may occur prior to the yielding of the ligament. Both models are suitable for thick-walled vessels. The paper discusses the applicability regime for both models. The models are then applied to a test vessel and the predicted failure pressure is compared against the pressure attained in the test. Results show that both models can be applied successfully. In particular, the contained yielding model when used with the plane-stress assumption can give reasonable predictions even for cases that involve yielding of the ligament. (orig.).

Integrity of PWR pressure vessels during overcooling accidents

International Nuclear Information System (INIS)

Cheverton, R.D.; Iskander, S.K.; Whitman, G.D.

1982-01-01

The reactor pressure vessel in a pressurized water reactor is normally subjected to temperatures and pressures that preclude propagation of sharp, crack-like defects that might exist in the wall of the vessel. However, there is a class of postulated accidents, referred to as overcooling accidents, that can subject the pressure vessel to severe thermal shock while the pressure is substantial. As a result of such accidents vessels containing high concentrations of copper and nickel, which enhance radiation embrittlement, may possess a potential for extensive propagation of preexistent inner surface flaws prior to the vessel's normal end of life. For the purpose of evaluating this problem a state-of-the-art fracture mechanics model was developed and has been used for conducting parametric analyses and for calculating several recorded PWR transients. Results of the latter analysis indicate that there may be some vessels that have a potential for failure today if subjected to a Rancho Seco (1978) or TMI-2 (1979) type transient. However, the calculational model may be excessively conservative, and this possibility is under investigation

Integrity of PWR pressure vessels during overcooling accidents

International Nuclear Information System (INIS)

Cheverton, R.D.; Iskander, S.K.; Whitman, G.D.

1982-01-01

The reactor pressure vessel in a pressurized water reactor is normally subjected to temperatures and pressures that preclude propagation of sharp, crack-like defects that might exist in the wall of the vessel. However, there is a class of postulated accidents, referred to as overcooling accidents, that can subject the pressure vessel to severe thermal shock while the pressure is substantial. As a result of such accidents, vessels containing high concentrations of copper and nickel, which enhance radiation embrittlement, may possess a potential for extensive propagation of preexistent inner surface flaws prior to the vessel's normal end of life. A state-of-the-art fracture-mechanics model was developed and has been used for conducting parametric analyses and for calculating several recorded PWR transients. Results of the latter analysis indicate that there may be some vessels that have a potential for failure in a few years if subjected to a Rancho Seco-type transient. However, the calculational model may be excessively conservative, and this possibility is under investigation

Numerical modeling of the pulse wave propagation in large blood vessels based on liquid and wall interaction

International Nuclear Information System (INIS)

Rup, K; Dróżdż, A

2014-01-01

The purpose of this article is to develop a non-linear, one-dimensional model of pulse wave propagation in the arterial cardiovascular system. The model includes partial differential equations resulting from the balance of mass and momentum for the fluid-filled area and the balance equation for the area of the wall and vessels. The considered mathematical model of pulse wave propagation in the thoracic aorta section takes into account the viscous dissipation of fluid energy, realistic values of parameters describing the physicochemical properties of blood and vessel wall. Boundary and initial conditions contain the appropriate information obtained from in vivo measurements. As a result of the numerical solution of the mass and momentum balance equations for the blood and the equilibrium equation for the arterial wall area, time- dependent deformation, respective velocity profiles and blood pressure were determined.

Analysis study on change of tendon behavior during pressurization process of Pre-stressed Concrete Containment Vessel

International Nuclear Information System (INIS)

Kashiwase, Takako; Nagasaka, Hideo

1999-01-01

NUPEC has been planning the ultimate strength test of Pre-stressed Concrete Containment Vessel (PCCV). The test model is 1/4 uniform scale model of Japan actual PCCV. It involves an equipment hatch, several penetrations and liner with T-anchors. The ancillary test for the PCCV test was conducted, in which friction coefficient of hoop tendon was evaluated by tensile force distribution using the same tendon as that of 1/4 PCCV model. Tendon will be in plastic region under internal pressure above 3.5 times design pressure (Pd) and surface characteristic of tendon and the resultant friction coefficient will be changed. In the present paper, tendon friction coefficient in the plastic region was obtained by evaluating plastic region data of tendon in the ancillary test. The validity of the obtained friction coefficient was confirmed by the tendon elongation data. In addition to the formally developed elastic region friction coefficient, the obtained plastic region correlation was incorporated into ABAQUS Ver. 5.6. The effect of tendon tensile force distribution change on structural behavior up to 3.8 Pd was evaluated. (author)

Modeling and measurement of the motion of the DIII-D vacuum vessel during vertical instabilities

International Nuclear Information System (INIS)

Reis, E.; Blevins, R.D.; Jensen, T.H.; Luxon, J.L.; Petersen, P.I.; Strait, E.J.

1991-11-01

The motions of the D3-D vacuum vessel during vertical instabilities of elongated plasmas have been measured and studied over the past five years. The currents flowing in the vessel wall and the plasma scrapeoff layer were also measured and correlated to a physics model. These results provide a time history load distribution on the vessel which were input to a dynamic analysis for correlation to the measured motions. The structural model of the vessel using the loads developed from the measured vessel currents showed that the calculated displacement history correlated well with the measured values. The dynamic analysis provides a good estimate of the stresses and the maximum allowable deflection of the vessel. In addition, the vessel motions produce acoustic emissions at 21 Hertz that are sufficiently loud to be felt as well as heard by the D3-D operators. Time history measurements of the sounds were correlated to the vessel displacements. An analytical model of an oscillating sphere provided a reasonable correlation to the amplitude of the measured sounds. The correlation of the theoretical and measured vessel currents, the dynamic measurements and analysis, and the acoustic measurements and analysis show that: (1) The physics model can predict vessel forces for selected values of plasma resistivity. The model also predicts poloidal and toroidal wall currents which agree with measured values; (2) The force-time history from the above model, used in conjunction with an axisymmetric structural model of the vessel, predicts vessel motions which agree well with measured values; (3) The above results, input to a simple acoustic model predicts the magnitude of sounds emitted from the vessel during disruptions which agree with acoustic measurements; (4) Correlation of measured vessel motions with structural analysis shows that a maximum vertical motion of the vessel up to 0.24 in will not overstress the vessel or its supports. 11 refs., 10 figs., 1 tab

Green vessel scheduling in liner shipping: Modeling carbon dioxide emission costs in sea and at ports of call

Directory of Open Access Journals (Sweden)

Maxim A. Dulebenets

2018-03-01

Full Text Available Considering a substantial increase in volumes of the international seaborne trade and drastic climate changes due to carbon dioxide emissions, liner shipping companies have to improve planning of their vessel schedules and improve energy efficiency. This paper presents a novel mixed integer non-linear mathematical model for the green vessel scheduling problem, which directly accounts for the carbon dioxide emission costs in sea and at ports of call. The original non-linear model is linearized and then solved using CPLEX. A set of numerical experiments are conducted for a real-life liner shipping route to reveal managerial insights that can be of importance to liner shipping companies. Results indicate that the proposed mathematical model can serve as an efficient planning tool for liner shipping companies and may assist with evaluation of various carbon dioxide taxation schemes. Increasing carbon dioxide tax may substantially change the design of vessel schedules, incur additional route service costs, and improve the environmental sustainability. However, the effects from increasing carbon dioxide tax on the marine container terminal operations are found to be very limited.

Ultimate Pressure Capacity of Prestressed Concrete Containment Vessels with Steel Fibers

Energy Technology Data Exchange (ETDEWEB)

Hahm, Dae Gi; Choun, Young Sun; Choi, In Kil [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2011-10-15

The ultimate pressure capacity (UPC) of the prestressed concrete containment vessel (PCCV) is very important since the PCCV are final protection to prevent the massive leakage of a radioactive contaminant caused by the severe accident of nuclear power plants (NPPs). The tensile behavior of a concrete is an important factor which influence to the UPC of PCCVs. Hence, nowadays, it is interested that the application of the steel fiber to the PCCVs since that the concrete with steel fiber shows an improved performance in the tensile behavior compared to reinforced concrete (RC). In this study, we performed the UPC analysis of PCCVs with steel fibers corresponding to the different volume ratio of fibers to verify the effectiveness of steel fibers on PCCVs

Improvement to reactor vessel

International Nuclear Information System (INIS)

1974-01-01

The vessel described includes a prestressed concrete vessel containing a chamber and a removable cover closing this chamber. The cover is in concrete and is kept in its closed position by main and auxiliary retainers, comprising fittings integral with the concrete of the vessel. The auxiliary retainers pass through the concrete of the cover. This improvement may be applied to BWR, PWR and LMFBR type reactor vessel [fr

Large panel design for containment air baffle

International Nuclear Information System (INIS)

Orr, R.S.

1992-01-01

The movable air baffle shield means in accordance with the present invention provides an efficient method of cooling the space surrounding the containment vessel while also providing the capability of being moved away from the containment vessel during inspection. The containment apparatus comprises a generally cylindrical sealed containment vessel for containing at least a portion of a nuclear power generation plant, a disparate shield building surrounding and housing the containment vessel therein and spaced outwardly thereof so as to form an air annulus in the space between the shield building and the containment vessel, a shield baffle means positioned in the air annulus around at least a portion of the sides of the containment vessel providing a coolant path between the baffle means and the containment vessel to permit cooling of the containment vessel by air, the shield baffle means being movable to afford access to the containment vessel. 9 figs

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)

Modelling of nonhomogeneous atmosphere in NPP containment using lumped-parameter model based on CFD calculations

International Nuclear Information System (INIS)

Kljenak, I.; Mavko, B.; Babic, M.

2005-01-01

Full text of publication follows: The modelling and simulation of atmosphere mixing and stratification in nuclear power plant containments is a topic, which is currently being intensely investigated. With the increase of computer power, it has now become possible to model these phenomena with a local instantaneous description, using so-called Computational Fluid Dynamics (CFD) codes. However, calculations with these codes still take relatively long times. An alternative faster approach, which is also being applied, is to model nonhomogeneous atmosphere with lumped-parameter codes by dividing larger control volumes into smaller volumes, in which conditions are modelled as homogeneous. The flow between smaller volumes is modelled using one-dimensional approaches, which includes the prescription of flow loss coefficients. However, some authors have questioned this approach, as it appears that atmosphere stratification may sometimes be well simulated only by adjusting flow loss coefficients to adequate 'artificial' values that are case-dependent. To start the resolution of this issue, a modelling of nonhomogeneous atmosphere with a lumped-parameter code is proposed, where the subdivision of a large volume into smaller volumes is based on results of CFD simulations. The basic idea is to use the results of a CFD simulation to define regions, in which the flow velocities have roughly the same direction. These regions are then modelled as control volumes in a lumped-parameter model. In the proposed work, this procedure was applied to a simulation of an experiment of atmosphere mixing and stratification, which was performed in the TOSQAN facility. The facility is located at the Institut de Radioprotection et de Surete Nucleaire (IRSN) in Saclay (France) and consists of a cylindrical vessel (volume: 7 m3), in which gases are injected. In the experiment, which was also proposed for the OECD/NEA International Standard Problem No.47, air was initially present in the vessel, and

A 1055 ft/sec impact test of a two foot diameter model nuclear reactor containment system without fracture

Science.gov (United States)

Puthoff, R. L.

1972-01-01

A study to determine the feasibility of containing the fission products of a mobile reactor in the event of an impact is presented. The model simulated the reactor core, energy absorbing gamma shielding, neutron shielding and the containment vessel. It was impacted against an 18,000 pound reinforced concrete block at 1055 ft/sec. The model was significantly deformed and the concrete block demolished. No leaks were detected nor were any cracks observed in the model after impact.

Polymer-based blood vessel models with micro-temperature sensors in EVE

Science.gov (United States)

Mizoshiri, Mizue; Ito, Yasuaki; Hayakawa, Takeshi; Maruyama, Hisataka; Sakurai, Junpei; Ikeda, Seiichi; Arai, Fumihito; Hata, Seiichi

2017-04-01

Cu-based micro-temperature sensors were directly fabricated on poly(dimethylsiloxane) (PDMS) blood vessel models in EVE using a combined process of spray coating and femtosecond laser reduction of CuO nanoparticles. CuO nanoparticle solution coated on a PDMS blood vessel model are thermally reduced and sintered by focused femtosecond laser pulses in atmosphere to write the sensors. After removing the non-irradiated CuO nanoparticles, Cu-based microtemperature sensors are formed. The sensors are thermistor-type ones whose temperature dependences of the resistance are used for measuring temperature inside the blood vessel model. This fabrication technique is useful for direct-writing of Cu-based microsensors and actuators on arbitrary nonplanar substrates.

Nuclear reactor containment device

International Nuclear Information System (INIS)

Ichiki, Tadaharu.

1980-01-01

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

Method for temporary shielding of reactor vessel internals

International Nuclear Information System (INIS)

Grimm, N.P.; Sejvar, J.

1991-01-01

This patent describes a method for shielding stored internals for reactor vessel annealing. It comprises removing nuclear fuel from the reactor vessel containment building; removing and storing upper and lower core internals under water in a refueling canal storage area; assembling a support structure in the refueling canal between the reactor vessel and the stored internals; introducing vertical shielding tanks individually through a hatch in the containment building and positioning each into the support structure; introducing horizontal shielding tanks individually through a hatch in the containment building and positioning each above the stored internals and vertical tanks; draining water from the refueling canal to the level of a flange of the reactor vessel; placing an annealing apparatus in the reactor vessel; pumping the remaining water from the reactor vessel; and annealing the reactor vessel

Hull Girder Fatigue Damage Estimations of a Large Container Vessel by Spectral Analysis

DEFF Research Database (Denmark)

Andersen, Ingrid Marie Vincent; Jensen, Jørgen Juncher

2013-01-01

This paper deals with fatigue damage estimation from the analysis of full-scale stress measurements in the hull of a large container vessel (9,400 TEU) covering several months of operation. For onboard decision support and hull monitoring sys-tems, there is a need for a fast reliable method...... for esti-mation of fatigue damage in the ship hull. The objective of the study is to investigate whether the higher frequency contributions from the hydroelastic responses (springing and whipping) can satisfactory be included in the fatigue damage estimation by only a few parameters derived from the stress...

Evaluation of temperature distribution in a containment vessel during operation

International Nuclear Information System (INIS)

Utanohara, Yoichi; Murase, Michio; Yanagi, Chihiro; Masui, Akihiro; Inomata, Ryo; Kamiya, Yuji

2012-01-01

For safety analysis of the containment vessel (CV) in a nuclear power plant, the average temperature of the gas phase in the CV during operation is used as an initial condition. An actual CV, however, has a temperature distribution, which makes the estimation of the average temperature difficult. Numerical simulation seems to be useful for the average temperature estimation, but it has several difficulties such as predictions of temperature distribution in a large and closed space that has several compartments, and modeling the heat generating components and the convection-diffusion of heat by ventilation air-conditioning systems. The main purpose of this study was to simulate the temperature distribution and evaluate the average temperature in the CV of a three-loop pressurized water reactor (PWR) during the reactor operation. The simulation considered the heat generation of equipment, flow due to the ventilation and air conditioning systems, heat loss to the CV exterior, and the solar heat. The predicted temperature distribution was significantly affected by the flow. Particularly, openings, which became flow paths, affected the temperature distribution. The temperature increased with a rise in height within the CV and the flow field seemed to transform from forced convection to natural convection. The volume-averaged temperature was different between gas and solid (concrete, CV wall) phases as well as between heights. The total volume-averaged temperature of the CV was nearly equal to the average gas phase temperature. It was found to be easy to evaluate the effect of openings on the temperature distribution and estimate the average temperature in CV by numerical simulation. (author)

Sargent and Lundy containment tests revisited

International Nuclear Information System (INIS)

Henry, Robert E.; Hammersley, Robert J.

2005-01-01

The pressurization experiments performed in the intermediate scale Sargent and Lundy containment test facility provide numerous insights into the dominant heat and mass transfer processes under design basis accident conditions similar to a large break Loss of Coolant Accident (LOCA). These experiments were the first integral tests to examine the containment response to a dynamic blowdown from the Reactor Coolant System (RCS). Measurements included the blowdown rate of the simulated Reactor Pressure Vessel (RPV), the pressure in containment as well as the containment temperatures in the top and bottom of the containment vessel. Furthermore, various experiments were performed with the blowdown location changed from the vessel bottom to the lower third of the vessel, the upper third of the vessel and near the top of the RPV to examine the influence of different types of break elevations, i.e. different characterizations of the exhausting steam-water mixture. Perhaps the most insightful set of measurements from these experiments were those that varied the cold water mass initially resident in the bottom of the simulated containment vessel. The role of this water as a function of its initial mass and the break location showed substantial influence of this water if the blowdown location provided sufficient energy to disperse this cold water into the containment building atmosphere. This is demonstrated in Figure 1 taken from Kolflat, 1960. All of these are relevant to an understanding of the dominant physical processes for this type of postulated accident condition. As such, it is important that all of these insights are retained and used in models for the containment building thermal-hydraulic response under accident conditions. Reference: Kolflat, A., 1960, 'Resulting of 1959 Nuclear Power Plant Containment Test', Sargent and Lundy Report SL-1800; Kolflat, A. and Chittenden, W. A., 1957, 'A New Approach to the Design of Containment Shells for Atomic Power Plants

Behavior of cracked concrete nuclear containment vessels during earthquakes

International Nuclear Information System (INIS)

Gergely, P.; Stanton, J.F.; White, R.N.

1975-01-01

When pressure builds up in a reinforced concrete nuclear containment shell, its cylindrical wall cracks vertically and horizontally at intervals of about five feet. If an earthquake occurs simultaneously with this pressurization, inertia forces are transmitted across the horizontal crack planes. The forces and deformations must be small enough to maintain the integrity of the steel liner. A typical containment shell has a radius of about 65 ft. and a wall thickness of about 4 ft. It is heavily reinforced with vertical, horizontal, and sometimes diagonal bars. A steel shell of about 3 / 8 in. thickness is attached to the concrete with anchors. The seismic shear forces are transmitted across the horizontal cracks by interface shear transfer (combination of shear friction and aggregate interlocking), by dowel action of the bars, and by diagonal bars if they are used. One important question in the design of such vessels is whether the diagonal bars are necessary. In the experimental portion of the current investigation several types of tests were conducted to study the load-slip characteristics of interface shear transfer under high intensity cyclic loading. In some cases external bars provided the clamping action of reinforcement, in more recent tests large diameter embedded bars were used. This presentation summarizes the analytical part of the investigation. A representative load-slip curve has been used in the analyses to assess the intensity of the stresses and deformations, and to study the importance of the variables as an aid in planning future tests

Nuclear reactors sited deep underground in steel containment vessels

Energy Technology Data Exchange (ETDEWEB)

Bourque, Robert [Los Alamos National Laboratory, P.O. Box 1663, Los Alamos, NM 87545 (United States)

2006-07-01

Although nuclear power plants are certainly very safe, they are not perceived as safe by the general populace. Also, there are concerns about overland transport of spent fuel rods and other irradiated components. It is hereby proposed that the nuclear components of nuclear power plants be placed in deep underground steel vessels with secondary coolant fed from them to turbines at or near the surface. All irradiated components, including spent fuel, would remain in the chamber indefinitely. This general concept was suggested by the late Edward Teller, generated some activity 20-25 years ago and appears to be recently reviving in interest. Previous work dealt with issues of geologic stability of underground, possibly reinforced, caverns. This paper presents another approach that makes siting independent of geology by placing the reactor components in a robust steel vessel capable of resisting full overburden pressure as well as pressures resulting from accident scenarios. Structural analysis of the two vessel concepts and approximate estimated costs are presented. This work clears the way for the extensive discussions required to evaluate the advantages of this concept. (author)

Dynamic analysis of the PEC fast reactor vessel: on-site tests and mathematical models

International Nuclear Information System (INIS)

Zola, Maurizio; Martelli, Alessandro; Maresca, Giuseppe; Masoni, Paolo; Scandola, Giani; Descleves, Pierre

1988-01-01

This paper presents the main features and results of the on-site dynamic tests and the related numerical analysis carried out for the PEC reactor vessel. The purpose is to provide an example of on-site testing of large components, stressing the problems encountered during the experiments, as well as in the processing phase of the test results and for the comparisons between calculations and measurements. Tests, performed by ISMES on behalf of ENEA, allowed the dynamic response of the empty vessel to be measured, thus providing data for the verification of the numerical models of the vessel supporting structure adopted in the PEC reactor-block seismic analysis. An axisymmetric model of the vessel, implemented in the vessel, implemented in the NOVAK code, had been developed in the framework of the detailed numerical studies performed by NOVATOME (again on behalf of ENEA), to check the beam schematization with fluid added mass model adopted by ANSALDO in SAP-IV and ANSYS for the reactor-block design calculations. Furthermore, a numerical model, describing vessel supporting structure in detail, was also developed by ANSALDO and implemented in the SAP-IV code. The test conditions were analysed by use of these and the design models. Comparisons between calculations and measurements showed particularly good agreement with regard to first natural frequency of the vessel and rocking stiffness of the vessel supporting structure, i.e. those parameters on which vessel seismic amplification mainly depends: this demonstrated the adequacy of the design analysis to correctly calculate the seismic motion at the PEC core diagrid. (author)

Numerical studies of large penetrations and closures for containment vessels subjected to loadings beyond the design basis

International Nuclear Information System (INIS)

Kulak, R.F.; Hsieh, B.J.; Kennedy, J.M.; Ash, J.E.; McLennan, G.A.

1984-01-01

Numerical simulations of the macro-deformations of the sealing surfaces (gasketed junctures) of a PWR steel containment vessel's equipment hatch and a BWR Mk II containment vessel head have been performed. Results for the equipment hatch juncture indicate that the rotations of the hatch cover and penetration sleeve must be accounted for when performing leakage analysis because they can effect the compression of the gasket even though the gasket is in a pressure-seated configuration. Results from a leakage analysis indicated that excessive leakage can occur if the surface roughness is high and/or the compression set is high. Results for the Mk II head show that both the temperature and pressure loadings must be taken into account to obtain realistic responses. The temperature difference between the flanges and bolts has the important net effect of keeping the gasketed juncture closed, that is in metal-to-metal contact. Due to the high accident temperature, the gasket itself was found to achieve 100% compression set and thus could not perform its sealing function within the juncture

Vessel size measurements in angiograms: A comparison of techniques

International Nuclear Information System (INIS)

Hoffmann, Kenneth R.; Nazareth, Daryl P.; Miskolczi, Laszlo; Gopal, Anant; Wang Zhou; Rudin, Stephen; Bednarek, Daniel R.

2002-01-01

As interventional procedures become more complicated, the need for accurate quantitative vascular information increases. In response to this need, many commercial vendors provide techniques for measurement of vessel sizes, usually based on derivative techniques. In this study, we investigate the accuracy of several techniques used in the measurement of vessel size. Simulated images of vessels having circular cross sections were generated and convolved with various focal spot distributions taking into account the magnification. These vessel images were then convolved with Gaussian image detector line spread functions (LSFs). Additionally, images of a phantom containing vessels with a range of diameters were acquired for the 4.5'', 6'', 9'', and 12'' modes of an image intensifier-TV (II-TV) system. Vessel sizes in the images were determined using a first-derivative technique, a second-derivative technique, a linear combination of these two measured sizes, a thresholding technique, a densitometric technique, and a model-based technique. For the same focal spot size, the shape of the focal spot distribution does not affect measured vessel sizes except at large magnifications. For vessels with diameters larger than the full-width-at-half-maximum (FWHM) of the LSF, accurate vessel sizes (errors ∼0.1 mm) could be obtained by using an average of sizes determined by the first and second derivatives. For vessels with diameters smaller than the FWHM of the LSF, the densitometric and model-based techniques can provide accurate vessel sizes when these techniques are properly calibrated

Stress categorization in nozzle to pressure vessel connections finite elements models

International Nuclear Information System (INIS)

Albuquerque, Levi Barcelos de

1999-01-01

The ASME Boiler and Pressure Vessel Code, Section III , is the most important code for nuclear pressure vessels design. Its design criteria were developed to preclude the various pressure vessel failure modes throughout the so-called 'Design by Analysis', some of them by imposing stress limits. Thus, failure modes such as plastic collapse, excessive plastic deformation and incremental plastic deformation under cyclic loading (ratchetting) may be avoided by limiting the so-called primary and secondary stresses. At the time 'Design by Analysis' was developed (early 60's) the main tool for pressure vessel design was the shell discontinuity analysis, in which the results were given in membrane and bending stress distributions along shell sections. From that time, the Finite Element Method (FEM) has had a growing use in pressure vessels design. In this case, the stress results are neither normally separated in membrane and bending stress nor classified in primary and secondary stresses. This process of stress separation and classification in Finite Element (FE) results is what is called stress categorization. In order to perform the stress categorization to check results from FE models against the ASME Code stress limits, mainly from 3D solid FE models, several research works have been conducted. This work is included in this effort. First, a description of the ASME Code design criteria is presented. After that, a brief description of how the FEM can be used in pressure vessel design is showed. Several studies found in the literature on stress categorization for pressure vessel FE models are reviewed and commented. Then, the analyses done in this work are presented in which some typical nozzle to pressure vessel connections subjected to internal pressure and concentrated loads were modeled with solid finite elements. The results from linear elastic and limit load analyses are compared to each other and also with the results obtained by formulae for simple shell

Reactor vessel supported by flexure member

International Nuclear Information System (INIS)

Crawford, J.D.; Pankow, B.

1977-01-01

According to the present invention there is provided an improved arrangement for supporting a reactor vessel within a containment structure against static and dynamic vertical loadings capable of being imposed as a result of a serious accident as well as during periods of normal plant operation. The support arrangement of the invention is, at the same time, capable of accommodating radial displacements that normally occur between the reactor vessel and the containment structure due to operational transients. The arrangement comprises a plurality of vertical columns connected between the reactor vessel and a support base within the containment structure. The columns are designed to accommodate relative displacements between the vessel and the containment structure by flexing. This eliminates the need for relative sliding movements and thus enables the columns to be securely fixed to the vessel. This elimination of a provision for relative sliding movements avoids the spaces or gaps between the retention members and the retained elements as occurred in prior art arrangements and, concomitantly, the danger of establishing impact forces on the retention members in the event of an accident is reduced. (author)

The TPX vacuum vessel and in-vessel components

International Nuclear Information System (INIS)

Heitzenroeder, P.; Bialek, J.; Ellis, R.; Kessel, C.; Liew, S.

1994-01-01

The Tokamak Physics Experiment (TPX) is a superconducting tokamak with double-null diverters. TPX is designed for 1,000-second discharges with the capability of being upgraded to steady state operation. High neutron yields resulting from the long duration discharges require that special consideration be given to materials and maintainability. A unique feature of the TPX is the use of a low activation, titanium alloy vacuum vessel. Double-wall vessel construction is used since it offers an efficient solution for shielding, bakeout and cooling. Contained within the vacuum vessel are the passive coil system, Plasma Facing Components (PFCs), magnetic diagnostics, and the internal control coils. All PFCs utilize carbon-carbon composites for exposed surfaces

Impact limiter design for a lightweight tritium hydride vessel transport container

International Nuclear Information System (INIS)

Harding, D.C.; Longcope, D.B.; Neilsen, M.K.

1995-01-01

Sandia National Laboratories (SNL) has designed an impact-limiting system for a small, lightweight radioactive material shipping container. The Westinghouse Savannah River Company (WSRC) is developing this Type B package for the shipment of tritium, replacing the outdated LP-50 shipping container. Regulatory accident resistance requirements for Type B packages, including this new tritium package, are specified in 10 CFR 71 (NRC 1983). The regulatory requirements include a 9-meter free drop onto an unyielding target, a 1-meter drop onto a mild steel punch, and a 30-minute 800 degrees C fire test. Impact limiters are used to protect the package in the free-drop accident condition in any impact orientation without hindering the package's resistance to the thermal accident condition. The overall design of the new package is based on a modular concept using separate thermal shielding and impact mitigating components in an attempt to simplify the design, analysis, test, and certification process. Performance requirements for the tritium package's limiting system are based on preliminary estimates provided by WSRC. The current tritium hydride vessel (THV) to be transported has relatively delicate valving assemblies and should not experience acceleration levels greater than approximately 200 g's. A thermal overpack and outer stainless steel shell, to be designed by WSRC, will form the inner boundary of the impact-limiting system (see Figure 1). The mass of the package, including cargo, inner container, thermal overpack, and outer stainless steel shell (not including impact limiters) should be approximately 68 kg. Consistent with the modular design philosophy, the combined thermal overpack and containment system should be considered essentially rigid, with the impact limiters incurring all deformation

Modelling of nonhomogeneous atmosphere in NPP containment using lumped-parameter model based on CFD calculations

International Nuclear Information System (INIS)

Ivo, Kljenak; Miroslav, Babic; Borut, Mavko

2007-01-01

The possibility of simulating adequately the flow circulation within a nuclear power plant containment using a lumped-parameter code is considered. An experiment on atmosphere mixing and stratification, which was performed in the containment experimental facility TOSQAN at IRSN (Institute of Radioprotection and Nuclear Safety) in Saclay (France), was simulated with the CFD (Computational Fluid Dynamics) code CFX4 and the lumped-parameter code CONTAIN. During some phases of the experiment, steady states were achieved by keeping the boundary conditions constant. Two steady states during which natural convection was the dominant gas flow mechanism were simulated independently. The nodalization of the lumped-parameter model was based on the flow pattern, simulated with the CFD code. The simulation with the lumped-parameter code predicted basically the same flow circulation patterns within the experimental vessel as the simulation with the CFD code did. (authors)

A 640 foot per second impact test of a two foot diameter model nuclear reactor containment system without fracture

Science.gov (United States)

Puthoff, R. L.

1971-01-01

An impact test was conducted on an 1142 pound 2 foot diameter sphere model. The purpose of this test was to determine the feasibility of containing the fission products of a mobile reactor in an impact. The model simulated the reactor core, energy absorbing gamma shielding, neutron shielding and the containment vessel. It was impacted against an 18,000 pound reinforced concrete block. The model was significantly deformed and the concrete block demolished. No leaks were detected nor cracks observed in the model after impact.

Finite element analysis of prestressed concrete reactor vessels

International Nuclear Information System (INIS)

Smith, P.D.; Cook, W.A.; Anderson, C.A.

1977-01-01

Several present and proposed gas-cooled reactors use concrete pressure vessels. In addition, concrete is almost universally used for the secondary containment structures of water-cooled reactors. Regulatory agencies must have means of assuring that these concrete structures perform their containment functions during normal operation and after extreme conditions of transient overpressure and high temperature. The NONSAP nonlinear structural analysis program has been extensively modified to provide one analytical means of assessing the safety of reinforced concrete pressure vessels and containments. Several structural analysis codes were studied to evaluate their ability to model the nonlinear static and dynamic behavior of three-dimensional structures. The NONSAP code was selected because of its availability and because of the ease with which it can be modified. In particular, the modular structure of this code allows ready addition of specialized material models. Major modifications have been the development of pre- and post-processors for mesh generation and graphics, the addition of an out-of-core solver, and the addition of constitutive models for reinforced concrete subject to either long-term or short-term loads. Emphasis was placed on development of a three-dimensional analysis capability

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.

Evaluation charts of thermal stresses in cylindrical vessels induced by thermal stratification of contained fluid

International Nuclear Information System (INIS)

Furuhashi, Ichiro; Kawasaki, Nobuchika; Kasahara, Naoto

2008-01-01

Temperature and thermal stress in cylindrical vessels were analysed for the thermal stratification of contained fluid. Two kinds of temperature analysis results were obtained such as the exact temperature solution of eigenfunction series and the simple approximate one by the temperature profile method. Furthermore, thermal stress shell solutions were obtained for the simple approximate temperatures. Through comparison with FEM analyses, these solutions were proved to be adequate. The simple temperature solution is described by one parameter that is the temperature decay coefficient. The thermal stress shell solutions are described by two parameters. One is the ratio between the temperature decay coefficient and the load decay coefficient. Another is the nondimensional width of stratification. These solutions are so described by few parameters that those are suitable for the simplified thermal stress evaluation charts. These charts enable quick and accurate thermal stress evaluations of cylindrical vessel of this problem compared with conventional methods. (author)

Evaluation charts of thermal stresses in cylindrical vessels induced by thermal stratification of contained fluid

International Nuclear Information System (INIS)

Furuhashi, Ichiro; Kawasaki, Nobuchika; Kasahara, Naoto

2007-01-01

Temperature and thermal stress in cylindrical vessels were analysed for the thermal stratification of contained fluid. Two kinds of temperature analysis results were obtained such as the exact temperature solution of eigen-function series and the simple approximate one by the temperature profile method. Furthermore, shell solutions of thermal stress were obtained for the simple approximate temperatures. Through comparison with FEM analyses, these solutions were proved to be adequate. The simple temperature solution is described by one parameter that is the temperature decay factor. The shell solutions of thermal stress are described by two parameters. One is the ratio between the temperature decay factor and the local decay factor. Another is the non-dimensional width of stratification. These solution are so described by few parameters that those are suitable for the simplified thermal stress evaluation charts. These charts enable quick and accurate thermal stress evaluations of cylindrical vessel of this problem compared with conventional methods. (author)

Containment severe accident thermohydraulic phenomena

International Nuclear Information System (INIS)

Frid, W.

1991-08-01

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

Dynamic analysis of the PEC fast reactor vessel: On-site tests and mathematical models

International Nuclear Information System (INIS)

Zola, M.; Martelli, A.; Masoni, P.; Scandola, G.

1988-01-01

This paper presents the main features and results of the on-site dynamic tests and the related numerical analyses carried out for the PEC reactor vessel. The purpose is to provide an example of on-site testing of large components, stressing the problems encountered during the experiments, as well as in the processing phase of the test results and for the comparisons between calculations and measurements. Tests, performed by ISMES on behalf of ENEA, allowed the dynamic response of the empty vessel to be measured, thus providing data for the verification of the numerical models of the vessel supporting structure adopted in the PEC reactor-block seismic analysis. An axisymmetric model of the vessel, implemented in the NOVAX code, had been developed in the framework of the detailed numerical studies performed by NOVATOME (again on behalf of ENEA), to check the beam schematization with fluid added mass model adopted by ANSALDO in SAP-IV and ANSYS for the reactor-block design calculations. Furthermore, a numerical model, describing vessel supporting structure in detail, was also developed by ANSALDO and implemented in the SAP-IV code. The test conditions were analysed by use of these and the design models. Comparisons between calculations and measurements showed particularly good agreement with regard to first natural frequency of the vessel and rocking stiffness of the vessel supporting structure, i.e. those parameters on which vessel seismic amplification mainly depends: this demonstrated the adequacy of the design analysis to correctly calculate the seismic motion at the PEC core diagrid. (author). 5 refs, 23 figs, 4 tabs

Segmentation and packaging reactor vessels internals

International Nuclear Information System (INIS)

Boucau, Joseph

2014-01-01

Document available in abstract form only, full text follows: With more than 25 years of experience in the development of reactor vessel internals and reactor vessel segmentation and packaging technology, Westinghouse has accumulated significant know-how in the reactor dismantling market. The primary challenges of a segmentation and packaging project are to separate the highly activated materials from the less-activated materials and package them into appropriate containers for disposal. Since disposal cost is a key factor, it is important to plan and optimize waste segmentation and packaging. The choice of the optimum cutting technology is also important for a successful project implementation and depends on some specific constraints. Detailed 3-D modeling is the basis for tooling design and provides invaluable support in determining the optimum strategy for component cutting and disposal in waste containers, taking account of the radiological and packaging constraints. The usual method is to start at the end of the process, by evaluating handling of the containers, the waste disposal requirements, what type and size of containers are available for the different disposal options, and working backwards to select a cutting method and finally the cut geometry required. The 3-D models can include intelligent data such as weight, center of gravity, curie content, etc, for each segmented piece, which is very useful when comparing various cutting, handling and packaging options. The detailed 3-D analyses and thorough characterization assessment can draw the attention to material potentially subject to clearance, either directly or after certain period of decay, to allow recycling and further disposal cost reduction. Westinghouse has developed a variety of special cutting and handling tools, support fixtures, service bridges, water filtration systems, video-monitoring systems and customized rigging, all of which are required for a successful reactor vessel internals

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

Dismantling id the reactor pressure vessel insulation and dissecting of the MZFR reactor pressure vessel

International Nuclear Information System (INIS)

Loeb, Andreas; Stanke, Dieter; Thoma, Markus; Eisenmann, Beata; Prechtl, Erwin; Dehnke, Burckhard

2008-01-01

The MZFR reactor was decommissioned in 1984. The authors describe the dismantling of the reactor pressure vessel insulation that consists of asbestos containing mineral fiber wool. The appropriate remote handling and cutting tools had to be adapted with respect to the restrained space in the containment. The dismantling of the reactor pressure vessel has been completed, the dissected parts have been packaged into 200 containers for the final repository Konrad. During the total project time no reportable events and no damage to persons occurred.

Dismantling method for reactor pressure vessel and system therefor

International Nuclear Information System (INIS)

Hayashi, Makoto; Enomoto, Kunio; Kurosawa, Koichi; Saito, Hideyo.

1994-01-01

Upon dismantling of a reactor pressure vessel, a containment building made of concretes is disposed underground and a spent pressure vessel is contained therein, and incore structures are contained in the spent pressure vessel. Further, a plasma-welder and a pressing machine are disposed to a pool for provisionally placing reactor equipments in the reactor building for devoluming the incore structures by welding and compression. An overhead-running crane and rails therefor are disposed on the roof and the outer side of the reactor building for transporting the pressure vessel from the reactor building to the containment building. They may be contained in the containment building after incorporation of the incore structures into the pressure vessel at the outside of the reactor building. For the devoluming treatment, a combination of cutting, welding, pressing and the like are optically conducted. A nuclear power plant can be installed by using a newly manufactured nuclear reactor, with no requirement for a new site and it is unnecessary to provide a new radioactive waste containing facility. (N.H.)

Theoretical modelling of physiologically stretched vessel in magnetisable stent assisted magnetic drug targeting application

International Nuclear Information System (INIS)

Mardinoglu, Adil; Cregg, P.J.; Murphy, Kieran; Curtin, Maurice; Prina-Mello, Adriele

2011-01-01

The magnetisable stent assisted magnetic targeted drug delivery system in a physiologically stretched vessel is considered theoretically. The changes in the mechanical behaviour of the vessel are analysed under the influence of mechanical forces generated by blood pressure. In this 2D mathematical model a ferromagnetic, coiled wire stent is implanted to aid collection of magnetic drug carrier particles in an elastic tube, which has similar mechanical properties to the blood vessel. A cyclic mechanical force is applied to the elastic tube to mimic the mechanical stress and strain of both the stent and vessel while in the body due to pulsatile blood circulation. The magnetic dipole-dipole and hydrodynamic interactions for multiple particles are included and agglomeration of particles is also modelled. The resulting collection efficiency of the mathematical model shows that the system performance can decrease by as much as 10% due to the effects of the pulsatile blood circulation. - Research highlights: →Theoretical modelling of magnetic drug targeting on a physiologically stretched stent-vessel system. →Cyclic mechanical force applied to mimic the mechanical stress and strain of both stent and vessel. →The magnetic dipole-dipole and hydrodynamic interactions for multiple particles is modelled. →Collection efficiency of the mathematical model is calculated for different physiological blood flow and magnetic field strength.

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

International Nuclear Information System (INIS)

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

1993-01-01

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

Improved nuclear reactor construction with bottom supported reactor vessel

International Nuclear Information System (INIS)

Sharbaugh, J.E.

1987-01-01

An improved liquid metal nuclear reactor construction has a reactor core and a generally cylindrical reactor vessel for holding liquid metal coolant and housing the core within the pool. A generally cylindrical concrete containment structure surrounds the reactor vessel and a central support pedestal is anchored to the containment structure base mat and supports the bottom wall of the reactor vessel and the reactor core. The periphery of the reactor vessel bore is supported by an annular structure which allows thermal expansion but not seismic motion of the vessel, and a bed of thermally insulating material uniformly supports the vessel base whilst allowing expansion thereof. A guard ring prevents lateral seismic motion of the upper end of the reactor vessel. The periphery of the core is supported by an annular structure supported by the vessel base and keyed to the vessel wall so as to be able to expand but not undergo seismic motion. A deck is supported on the containment structure above the reactor vessel open top by annular bellows, the deck carrying the reactor control rods such that heating of the reactor vessel results in upward expansion against the control rods. (author)

Computed tomography depiction of small pediatric vessels with model-based iterative reconstruction

Energy Technology Data Exchange (ETDEWEB)

Koc, Gonca; Courtier, Jesse L.; Phelps, Andrew; Marcovici, Peter A.; MacKenzie, John D. [UCSF Benioff Children' s Hospital, Department of Radiology and Biomedical Imaging, San Francisco, CA (United States)

2014-07-15

Computed tomography (CT) is extremely important in characterizing blood vessel anatomy and vascular lesions in children. Recent advances in CT reconstruction technology hold promise for improved image quality and also reductions in radiation dose. This report evaluates potential improvements in image quality for the depiction of small pediatric vessels with model-based iterative reconstruction (Veo trademark), a technique developed to improve image quality and reduce noise. To evaluate Veo trademark as an improved method when compared to adaptive statistical iterative reconstruction (ASIR trademark) for the depiction of small vessels on pediatric CT. Seventeen patients (mean age: 3.4 years, range: 2 days to 10.0 years; 6 girls, 11 boys) underwent contrast-enhanced CT examinations of the chest and abdomen in this HIPAA compliant and institutional review board approved study. Raw data were reconstructed into separate image datasets using Veo trademark and ASIR trademark algorithms (GE Medical Systems, Milwaukee, WI). Four blinded radiologists subjectively evaluated image quality. The pulmonary, hepatic, splenic and renal arteries were evaluated for the length and number of branches depicted. Datasets were compared with parametric and non-parametric statistical tests. Readers stated a preference for Veo trademark over ASIR trademark images when subjectively evaluating image quality criteria for vessel definition, image noise and resolution of small anatomical structures. The mean image noise in the aorta and fat was significantly less for Veo trademark vs. ASIR trademark reconstructed images. Quantitative measurements of mean vessel lengths and number of branches vessels delineated were significantly different for Veo trademark and ASIR trademark images. Veo trademark consistently showed more of the vessel anatomy: longer vessel length and more branching vessels. When compared to the more established adaptive statistical iterative reconstruction algorithm, model

Computed tomography depiction of small pediatric vessels with model-based iterative reconstruction

International Nuclear Information System (INIS)

Koc, Gonca; Courtier, Jesse L.; Phelps, Andrew; Marcovici, Peter A.; MacKenzie, John D.

2014-01-01

Computed tomography (CT) is extremely important in characterizing blood vessel anatomy and vascular lesions in children. Recent advances in CT reconstruction technology hold promise for improved image quality and also reductions in radiation dose. This report evaluates potential improvements in image quality for the depiction of small pediatric vessels with model-based iterative reconstruction (Veo trademark), a technique developed to improve image quality and reduce noise. To evaluate Veo trademark as an improved method when compared to adaptive statistical iterative reconstruction (ASIR trademark) for the depiction of small vessels on pediatric CT. Seventeen patients (mean age: 3.4 years, range: 2 days to 10.0 years; 6 girls, 11 boys) underwent contrast-enhanced CT examinations of the chest and abdomen in this HIPAA compliant and institutional review board approved study. Raw data were reconstructed into separate image datasets using Veo trademark and ASIR trademark algorithms (GE Medical Systems, Milwaukee, WI). Four blinded radiologists subjectively evaluated image quality. The pulmonary, hepatic, splenic and renal arteries were evaluated for the length and number of branches depicted. Datasets were compared with parametric and non-parametric statistical tests. Readers stated a preference for Veo trademark over ASIR trademark images when subjectively evaluating image quality criteria for vessel definition, image noise and resolution of small anatomical structures. The mean image noise in the aorta and fat was significantly less for Veo trademark vs. ASIR trademark reconstructed images. Quantitative measurements of mean vessel lengths and number of branches vessels delineated were significantly different for Veo trademark and ASIR trademark images. Veo trademark consistently showed more of the vessel anatomy: longer vessel length and more branching vessels. When compared to the more established adaptive statistical iterative reconstruction algorithm, model

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

Study on the application of 50 mm thick welded joints without PWHT for containment vessels

Energy Technology Data Exchange (ETDEWEB)

Watanabe, Nozomu; Sakai, Yoshiyuki; Hayashi, Kazutoshi; Higashikubo, Tomohiro (Mitsubishi Heavy Industries. Ltd., Kobe Shipyard and Machinery Works (Japan)); Iida, Kunihiro (Shibaura Inst. of Tech., Dept. of Mechanical Engineering, Tokyo (Japan)); Satou, Masanobu (Mitsubishi Heavy Industries. Ltd., Tkasago Research and Development Center (Japan))

1992-01-01

In order to investigate the propriety of the use of 50 mm thick SGV480 carbon steel which is equivalent to ASTM A516 Gr. 70 without post weld heat treatment for containment vessels, the authors have certified the basic properties of base metal and welded joints of 50 mm thick SGV480 steel plates. The results showed that fracture thoughness of welded joints is high without PWHT and the steel is safe enough without PWHT against embrittlement fracture under the operating conditions. (orig.).

Study on the application of 50 mm thick welded joints without PWHT for containment vessels

International Nuclear Information System (INIS)

Watanabe, Nozomu; Sakai, Yoshiyuki; Hayashi, Kazutoshi; Higashikubo, Tomohiro; Iida, Kunihiro; Satou, Masanobu

1992-01-01

In order to investigate the propriety of the use of 50 mm thick SGV480 carbon steel which is equivalent to ASTM A516 Gr. 70 without post weld heat treatment for containment vessels, the authors have certified the basic properties of base metal and welded joints of 50 mm thick SGV480 steel plates. The results showed that fracture thoughness of welded joints is high without PWHT and the steel is safe enough without PWHT against embrittlement fracture under the operating conditions. (orig.)

The Container Stowage Problem

DEFF Research Database (Denmark)

Janstrup, Kira; Rose, Trine Høyer; Andersen, Kent Høj

The main purpose of this project is to use integer programming to create a model that minimizes the costs for container transportation by ship. To make the model as realistic as possible it will be based on information from a large shipping company about the vessel layout and container types....... In addition to our project two other projects are made where an optimal solution to the container stowage problem also is tried to be found, but by using constraint programming and local search instead respectively. We will therefore in the end compare these three methods and the achieved results on fastness...

The Container Stowage Problem

DEFF Research Database (Denmark)

Janstrup, Kira

2010-01-01

The main purpose of this project is to use integer programming to create a model that minimizes the costs for container transportation by ship. To make the model as realistic as possible it will be based on information from a large shipping company about the vessel layout and container types....... In addition to our project two other projects are made where an optimal solution to the container stowage problem also is tried to be found, but by using constraint programming and local search instead respectively. We will therefore in the end compare these three methods and the achieved results on fastness...

Development of ultrasonic testing technique with the large transducer to inspect the containment vessel plates of nuclear power plant embedded in concrete

International Nuclear Information System (INIS)

Ishida, Hitoshi; Kurozumi, Yasuo; Kaneshima, Yoshiari

2004-01-01

The containment vessel plates embedded in concrete on Pressurized Water Reactors are inaccessible to inspect directly. Therefore, it is advisable to prepare inspection technology to detect existence and a location of corrosion on the embedded plates indirectly. In order to establish ultrasonic testing technique to be able to inspect the containment vessel plates embedded in concrete widely at the accessible point, experiments to detect artificial hollows simulating corrosion on a surface of a carbon steel plate mock-up covered with concrete simulating the embedded containment vessel plates were carried out with newly made ultrasonic transducers. We made newly low frequency (0.3 MHz and 0.5 MHz) surface shear horizontal (SH) wave transducers combined with three large active elements, which were equivalent to a 120mm width element. As a result of the experiments, the surface SH transducers could detect clearly the echo from the hollows with a depth of 9.5 mm and 19 mm at a distance of 1500mm from the transducers on the surface of the mock-up covered with concrete. Therefore, we evaluate that it is possible to detect the defects such as corrosion on the plates embedded in concrete with the newly made low frequency surface SH transducers with large elements. (author)

Dynamic Properties of Container Vessel with Low Metacentric Height

DEFF Research Database (Denmark)

Blanke, M.; Jensen, A.G.

1997-01-01

between roll and lateral motions. This was changed with the construction of a unique four degrees of freedom roll planar motion mechanism (RPMM) at the Danish Maritime Institute. The paper presents complete nonlinear models for a container ship obtained with this facility. Model scale predictions...

Reactor containment

International Nuclear Information System (INIS)

Kawabe, Ryuhei; Yamaki, Rika.

1990-01-01

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

Rebound coefficient of collisionless gas in a rigid vessel. A model of reflection of field-reversed configuration

International Nuclear Information System (INIS)

Takaku, Yuichi; Hamada, Shigeo

1996-01-01

A system of collisionless neutral gas contained in a rigid vessel is considered as a simple model of reflection of field-reversed configuration (FRC) plasma by a magnetic mirror. The rebound coefficient of the system is calculated as a function of the incident speed of the vessel normalized by the thermal velocity of the gas before reflection. The coefficient is compared with experimental data of FIX (Osaka U.) and FRX-C/T(Los Alamos N.L.). Agreement is good for this simple model. Interesting is that the rebound coefficient takes the smallest value (∼0.365) as the incident speed tends to zero and approaches unity as it tends to infinity. This behavior is reverse to that expected for a system with collision dominated fluid instead of collisionless gas. By examining the rebound coefficient, therefore, it could be successfully inferred whether the ion mean free path in each experiment was longer or shorter than the plasma length. (author)

FFTF and CRBRP reactor vessels

International Nuclear Information System (INIS)

Morgan, R.E.

1977-01-01

The Fast Flux Test Facility (FFTF) reactor vessel and the Clinch River Breeder Reactor Plant (CRBRP) reactor vessel each serve to enclose a fast spectrum reactor core, contain the sodium coolant, and provide support and positioning for the closure head and internal structure. Each vessel is located in its reactor cavity and is protected by a guard vessel which would ensure continued decay heat removal capability should a major system leak develop. Although the two plants have significantly different thermal power ratings, 400 megawatts for FFTF and 975 megawatts for CRBRP, the two reactor vessels are comparable in size, the CRBRP vessel being approximately 28% longer than the FFTF vessel. The FFTF vessel diameter was controlled by the space required for the three individual In-Vessel Handling Machines and Instrument Trees. Utilization of the triple rotating plug scheme for CRBRP refueling enables packaging of the larger CRBRP core in a vessel the same diameter as the FFTF vessel

Targeting Therapy Resistant Tumor Vessels

Science.gov (United States)

2008-08-01

Morris LS. Hysterectomy vs. resectoscopic endometrial ablation for the control of abnormal uterine bleeding . A cost-comparative study. J Reprod Med 1994;39...after the antibody treatment contain a pericyte coat, vessel architecture is normal, the diameter of the vessels is smaller (dilated, abnormal vessels...involvement of proteases from inflammatory mast cells and functionally abnormal (Carmeliet and Jain, 2000; Pasqualini (Coussens et al., 1999) and other bone

Continuum mathematical modelling of pathological growth of blood vessels

Science.gov (United States)

Stadnik, N. E.; Dats, E. P.

2018-04-01

The present study is devoted to the mathematical modelling of a human blood vessel pathological growth. The vessels are simulated as the thin-walled circular tube. The boundary value problem of the surface growth of an elastic thin-walled cylinder is solved. The analytical solution is obtained in terms of velocities of stress strain state parameters. The condition of thinness allows us to study finite displacements of cylinder surfaces by means of infinitesimal deformations. The stress-strain state characteristics, which depend on the mechanical parameters of the biological processes, are numerically computed and graphically analysed.

Threshold for sweepout from pedestal region of Mark III containment

International Nuclear Information System (INIS)

Sienicki, J.J.; Spencer, B.W.

1984-01-01

The assessment of the consequences of highly unlikely severe accident sequences in boiling water reactors includes those sequences in which molten corium is postulated to meltthrough the reactor pressure vessel (RPV) lower head and enter the pedestal region beneath the vessel. If localized melt-through of the reactor vessel occurs at elevated primary system pressure, the ejection of molten corium from the vessel will be followed by a blowdown of steam and hydrogen. The gases flowing from the breached vessel constitute a source of driving forces capable of dispersing corium from the pedestal into other parts of the containment. The extent of the gas blowdown-driven sweepout process depends upon a number of factors including the primary system pressure at melt through, breach flow area, overall blowdown timescale, and the specific pedestal/containment geometry. A model is presented to predict whether or not the conditions of gas flow from the failed RPV are sufficient to cause sweepout of corium and/or water from the pedestal. The model is shown to predict the onset of sweepout in scale model, simulant material experiments and is applied to the investigation of sweepout in the full-size reactor system

A model for ultrasound contrast agent in a phantom vessel

KAUST Repository

Qamar, Adnan; Samtaney, Ravi

2014-01-01

A theoretical framework to model the dynamics of Ultrasound Contrast Agent (UCA) inside a phantom vessel is presented. The model is derived from the reduced Navier-Stokes equation and is coupled with the evolving flow field solution inside

Radioactive waste processing vessel

International Nuclear Information System (INIS)

Hayashi, Masaru; Suzuki, Osamu; Ishizaki, Kanjiro.

1987-01-01

Purpose: To obtain a vessel of a reduced weight and with no external leaching of radioactive materials. Constitution: The vessel main body is constituted, for example, with light weight concretes or foamed concretes, particularly, foamed concretes containing fine closed bubbles in the inside. Then, layers having dense texture made of synthetic resin such as polystylene, vinylchloride resin, etc. or metal plate such as stainless plate are integrally disposed to the inner surface of the vessel main body. The cover member also has the same structure. (Sekiya, K.)

Analyses of containment structures with corrosion damage

International Nuclear Information System (INIS)

Cherry, J.L.

1997-01-01

Corrosion damage that has been found in a number of nuclear power plant containment structures can degrade the pressure capacity of the vessel. This has prompted concerns regarding the capacity of corroded containments to withstand accident loadings. To address these concerns, finite element analyses have been performed for a typical PWR Ice Condenser containment structure. Using ABAQUS, the pressure capacity was calculated for a typical vessel with no corrosion damage. Multiple analyses were then performed with the location of the corrosion and the amount of corrosion varied in each analysis. Using a strain-based failure criterion, a open-quotes lower boundclose quotes, open-quotes best estimateclose quotes, and open-quotes upper boundclose quotes failure level was predicted for each case. These limits were established by: determining the amount of variability that exists in material properties of typical containments, estimating the amount of uncertainty associated with the level of modeling detail and modeling assumptions, and estimating the effect of corrosion on the material properties

Buffer lining manufacturing method for radioactive waste container

International Nuclear Information System (INIS)

Kawakami, Susumu; Sugino, Hiroyuki

1998-01-01

A recessed portion is formed on an upper surface of a filler layer made of a buffer powder filled into a container main body, the upper portion of the vessel main body is closed by a shrinkable liquid tight film. It is placed in a pressurizing container and pressed to mold a buffer lining base material integrated with the vessel main body. A flat upper surface and a containing space are formed by shaving to form a buffer lining. A disposing vessel containing radioactive wastes is inserted into the containing space, and the containing space is closed by a buffer block. The upper surface is sealed by a lid. With such a constitution, since a buffer lining integrated with the vessel main body can be formed easily inside the vessel main body, the disposing vessel can be contained in the containing vessel in a state surrounded by the buffer easily and stably without laying or piling over a large quantity of buffer blocks. (T.M.)

Development of a surrogate model for analysis of ex-vessel steam explosion in Nordic type BWRs

Energy Technology Data Exchange (ETDEWEB)

Grishchenko, Dmitry, E-mail: dmitry@safety.sci.kth.se; Basso, Simone, E-mail: simoneb@kth.se; Kudinov, Pavel, E-mail: pavel@safety.sci.kth.se

2016-12-15

Highlights: • Severe accident. • Steam explosion. • Surrogate model. • Sensitivity study. • Artificial neural networks. - Abstract: Severe accident mitigation strategy adopted in Nordic type Boiling Water Reactors (BWRs) employs ex-vessel core melt cooling in a deep pool of water below reactor vessel. Energetic fuel–coolant interaction (steam explosion) can occur during molten core release into water. Dynamic loads can threaten containment integrity increasing the risk of fission products release to the environment. Comprehensive uncertainty analysis is necessary in order to assess the risks. Computational costs of the existing fuel–coolant interaction (FCI) codes is often prohibitive for addressing the uncertainties, including the effect of stochastic triggering time. This paper discusses development of a computationally efficient surrogate model (SM) for prediction of statistical characteristics of steam explosion impulses in Nordic BWRs. The TEXAS-V code was used as the Full Model (FM) for the calculation of explosion impulses. The surrogate model was developed using artificial neural networks (ANNs) and the database of FM solutions. Statistical analysis was employed in order to treat chaotic response of steam explosion impulse to variations in the triggering time. Details of the FM and SM implementation and their verification are discussed in the paper.

Optimization of Container Line Networks with Flexible Demands

DEFF Research Database (Denmark)

Plum, Christian Edinger Munk

on real world problem instances of 500+ vessels and 500+ contracts. The method allows a global liner carrier to efficiently plan bunker purchases for their vessels, using a large number of bunker contracts to lower costs. Container vessels operate on tight schedules to meet the customers transit time...... is investigated and models supporting decision making is developed, with the aim of reducing costs, bunker consumption and increasing the revenue and service levels of a liner shipping company. These problems are complex, dealing with millions of containers traveling on hundreds of vessels calling hundreds...... aspects where important and relevant to include. To alleviate this a thorough description of the domain of liner shipping is given, explaining the industry in the words of an operations researcher. At the same time a set of benchmark instances LINER-LIB 2012 is introduced. It is the hope...

Direct containment heating models in the CONTAIN code

International Nuclear Information System (INIS)

Washington, K.E.; Williams, D.C.

1995-08-01

The potential exists in a nuclear reactor core melt severe accident for molten core debris to be dispersed under high pressure into the containment building. If this occurs, the set of phenomena that result in the transfer of energy to the containment atmosphere and its surroundings is referred to as direct containment heating (DCH). Because of the potential for DCH to lead to early containment failure, the U.S. Nuclear Regulatory Commission (USNRC) has sponsored an extensive research program consisting of experimental, analytical, and risk integration components. An important element of the analytical research has been the development and assessment of direct containment heating models in the CONTAIN code. This report documents the DCH models in the CONTAIN code. DCH models in CONTAIN for representing debris transport, trapping, chemical reactions, and heat transfer from debris to the containment atmosphere and surroundings are described. The descriptions include the governing equations and input instructions in CONTAIN unique to performing DCH calculations. Modifications made to the combustion models in CONTAIN for representing the combustion of DCH-produced and pre-existing hydrogen under DCH conditions are also described. Input table options for representing the discharge of debris from the RPV and the entrainment phase of the DCH process are also described. A sample calculation is presented to demonstrate the functionality of the models. The results show that reasonable behavior is obtained when the models are used to predict the sixth Zion geometry integral effects test at 1/10th scale

Direct containment heating models in the CONTAIN code

Energy Technology Data Exchange (ETDEWEB)

Washington, K.E.; Williams, D.C.

1995-08-01

The potential exists in a nuclear reactor core melt severe accident for molten core debris to be dispersed under high pressure into the containment building. If this occurs, the set of phenomena that result in the transfer of energy to the containment atmosphere and its surroundings is referred to as direct containment heating (DCH). Because of the potential for DCH to lead to early containment failure, the U.S. Nuclear Regulatory Commission (USNRC) has sponsored an extensive research program consisting of experimental, analytical, and risk integration components. An important element of the analytical research has been the development and assessment of direct containment heating models in the CONTAIN code. This report documents the DCH models in the CONTAIN code. DCH models in CONTAIN for representing debris transport, trapping, chemical reactions, and heat transfer from debris to the containment atmosphere and surroundings are described. The descriptions include the governing equations and input instructions in CONTAIN unique to performing DCH calculations. Modifications made to the combustion models in CONTAIN for representing the combustion of DCH-produced and pre-existing hydrogen under DCH conditions are also described. Input table options for representing the discharge of debris from the RPV and the entrainment phase of the DCH process are also described. A sample calculation is presented to demonstrate the functionality of the models. The results show that reasonable behavior is obtained when the models are used to predict the sixth Zion geometry integral effects test at 1/10th scale.

Improved Wave-vessel Transfer Functions by Uncertainty Modelling

DEFF Research Database (Denmark)

Nielsen, Ulrik Dam; Fønss Bach, Kasper; Iseki, Toshio

2016-01-01

This paper deals with uncertainty modelling of wave-vessel transfer functions used to calculate or predict wave-induced responses of a ship in a seaway. Although transfer functions, in theory, can be calculated to exactly reflect the behaviour of the ship when exposed to waves, uncertainty in inp...

An 810 ft/sec soil impact test of a 2-foot diameter model nuclear reactor containment system

Science.gov (United States)

Puthoff, R. L.

1972-01-01

A soil impact test was conducted on a 880-pound 2-foot diameter sphere model. The impact area consisted of back filled desert earth and rock. The impact generated a crater 5 feet in diameter by 5 feet deep. It buried itself a total of 15 feet - as measured to the bottom of the model. After impact the containment vessel was pressure checked. No leaks were detected nor cracks observed.

Physical modelling of LNG rollover in a depressurized container filled with water

Science.gov (United States)

Maksim, Dadonau; Denissenko, Petr; Hubert, Antoine; Dembele, Siaka; Wen, Jennifer

2015-11-01

Stable density stratification of multi-component Liquefied Natural Gas causes it to form distinct layers, with upper layer having a higher fraction of the lighter components. Heat flux through the walls and base of the container results in buoyancy-driven convection accompanied by heat and mass transfer between the layers. The equilibration of densities of the top and bottom layers, normally caused by the preferential evaporation of Nitrogen, may induce an imbalance in the system and trigger a rapid mixing process, so-called rollover. Numerical simulation of the rollover is complicated and codes require validation. Physical modelling of the phenomenon has been performed in a water-filled depressurized vessel. Reducing gas pressure in the container to levels comparable to the hydrostatic pressure in the water column allows modelling of tens of meters industrial reservoirs using a 20 cm laboratory setup. Additionally, it allows to model superheating of the base fluid layer at temperatures close the room temperature. Flow visualizations and parametric studies are presented. Results are related to outcomes of numerical modelling.

Neutron Assay System for Con?nement Vessel Disposition

International Nuclear Information System (INIS)

Frame, Katherine C.; Bourne, Mark M.; Crooks, William J.; Evans, Louise; Mayo, Douglas R.; Miko, David K.; Salazar, William R.; Stange, Sy; Valdez, Jose I.; Vigil, Georgiana M.

2012-01-01

Waste will be removed from confinement vessels remaining from 1970s-era experiments. Los Alamos has 9+ spherical confinement vessels remaining from experiments. Each vessel contains ∼ 500 lbs of radioactive debris such as actinide metals and oxides, metals, powdered silica, graphite, and wires and hardware. In order to dispose of the vessels, debris and contamination must be removed. Neutron assay system was designed to assay vessels before and after cleanout. System requirements are: (1) Modular and moveable; (2) Capable of detecting ∼100g 239 Pu equivalent in a 2-inch thick steel sphere with 6 foot diameter; and (3) Capable of safeguards-quality assays. Initial design parameters arethe use of 4-atm 3 He tubes with length of 6 feet, and 3 He tubes embedded in polyethelene for moderation. This paper describes the calibration of the Confinement Vessel Assay System (CVAS) and quantification of its uncertainties. Assay uncertainty depends on five factors: (1) Statistical uncertainty in the assay measurement; (2) Statistical uncertainty in the background measurement; (3) Statistical uncertainty in the isotopics determination - This should be much smaller than the other uncertainties; (4) Systematic uncertainty due to position bias; and (5) Systematic uncertainty due to fluctuations in cosmic ray spallation. This one can be virtually eliminated by performing the background measurement with an empty vessel - but that may not be possible. We used modeling and experiments to quantify the systematic uncertainties. The calibration assumes a uniform distribution of material, but reality will be different. MCNPX modeling was used to quantify the positional bias. The model was benchmarked to build confidence in its results. Material at top of vessel is 44% greater than amount assayed, according to singles. Material near 19-tube detector is 38% less than amount assayed, according to singles. Cosmic ray spallation contributes significantly to the background. Comparing rates

Nuclear Power Plant Prestressed Concrete Containment Vessel Structure Monitoring during Integrated Leakage Rate Testing Using Fiber Bragg Grating Sensors

Directory of Open Access Journals (Sweden)

Jinke Li

2017-04-01

Full Text Available As the last barrier of nuclear reactor, prestressed concrete containment vessels (PCCVs play an important role in nuclear power plants (NPPs. To test the mechanical property of PCCV during the integrated leakage rate testing (ILRT, a fiber Bragg grating (FBG sensor was used to monitor concrete strain. In addition, a finite element method (FEM model was built to simulate the progress of the ILRT. The results showed that the strain monitored by FBG had the same trend compared to the inner pressure variation. The calculation results showed a similar trend compared with the monitoring results and provided much information about the locations in which the strain sensors should be installed. Therefore, it is confirmed that FBG sensors and FEM simulation are very useful in PCCV structure monitoring.

Scatter modelling of fracture toughness data for reactor pressure vessel structural integrity assessment

International Nuclear Information System (INIS)

Pesoz, M.

1997-01-01

In the last decade, there has been an increasing interest at EDF in developing and applying probabilistic methods for a variety of purposes. In the field of structural integrity and reliability they are used to evaluate the effect of deterioration due to ageing mechanisms, mainly on major passive structural components such as reactor pressure vessel, steam generator and piping in nuclear plants. Such approaches provide an attractive supplement to the more conventional deterministic method, based upon pessimistic assumptions, that give results too far from reality to support effective decisions. In addition to deterministic calculations, a Probabilistic Fracture Mechanics model has been developed in order to analyse the risk of brittle failure of the reactor pressure vessel and to perform sensitivity studies. The material fracture toughness (K IC ) uncertainty appears to be strongly influencing the probability of failure under accidental conditions. Up to now, this parameter is determined from the RCC-M code reference curve, which is the same as the ASME reference curve. But an important issue when performing probabilistic analysis is the correct statistical modelling of input parameters. That's why modelling works have been carried out using results of fracture toughness tests performed for demonstrating the validity of the reference curve. This paper presents the statistical treatments that have been performed to model the scatter of temperature dependent parameter (K IC (T). A specific data base containing a few hundreds of French and US results have been carried and Weibull models have been fitted, based on various master curve equations (K. Wallin (Senior Adviser at the Technical Research Centre of Finland) or RCC-M types). (author)

Probabilistic Assessment of the Design and Safety of HSLA-100 Steel Confinement Vessels

Energy Technology Data Exchange (ETDEWEB)

R.M. Dolin

2003-03-03

This probabilistic approach for assessing the design and safety of the HSLA-100 steel confinement vessel used for a DynEx test involved the probability of failure for several scenarios, in which a fragment may penetrate the vessel. The samples involve vessel thicknesses of 1 inch, 2 inches, and 5.25 inches--the combined thicknesses of the 2 inch containment vessel and the 3.25 inch safety vessel. Two simulation approaches were used for each scenario to assess the probability of failure. The Likelihood of Occurrence method simultaneously models all likely fragment events of a test, for which the net probability of failure is the sum of all the fragment events. The Stochastic Sampling method determines the probability of a fragment perforation on the basis of a logical model and takes the overall probability that an experiment results in failure as the maximum probability for any fragment event. With margin and safety assessments taken into account, it was concluded that the one and two inch thicknesses by themselves are inadequate for containing a DynEx test. The 5.25 inch thickness was determined to be safe by the Likelihood of Occurrence method and nearly adequate by the Stochastic Sampling simulation.

Establishment of an animal model of mice with radiation- injured soft tissue blood vessels

International Nuclear Information System (INIS)

Wang Daiyou; Yu Dahai; Wu Jiaxiao; Wei Shanliang; Wen Yuming

2004-01-01

Objective: The aim of this study was to establish an animal model of mice with radiation-injured soft tissue blood vessels. Methods: Forty male mice were irradiated with 30 Gy on the right leg. After the irradiation was finished each of the 40 male mice was tested with angiography, and its muscle tissues on the bilateral legs were examined with vessel staining assay and electron microscopy. Results: The results showed that the number of vessels on the right leg was less than that on the left leg, the microvessel density, average diameter and average sectional area of the right leg were all lower than those of the left, and the configuration and ultra-structure of vessels were also different between both sides of legs. Conclusion: In the study authors successfully established an animal model of mice with radiation-injured soft tissue blood vessels

Intravenous injection of artificial red cells and subsequent dye laser irradiation causes deep vessel impairment in an animal model of port-wine stain.

Science.gov (United States)

Rikihisa, Naoaki; Tominaga, Mai; Watanabe, Shoji; Mitsukawa, Nobuyuki; Saito, Yoshiaki; Sakai, Hiromi

2018-03-15

Our previous study proposed using artificial blood cells (hemoglobin vesicles, Hb-Vs) as photosensitizers in dye laser treatment for port-wine stains (PWSs). Dye laser photons are absorbed by red blood cells (RBCs) and hemoglobin (Hb) mixture, which potentially produce more heat and photocoagulation and effectively destroy endothelial cells. Hb-Vs combination therapy will improve clinical outcomes of dye laser treatment for PWSs because very small vessels do not contain sufficient RBCs and they are poor absorbers/heaters of lasers. In the present study, we analyzed the relationship between vessel depth from the skin surface and vessel distraction through dye laser irradiation following intravenous Hb-Vs injection using a chicken wattle model. Hb-Vs were administered and chicken wattles underwent high-energy irradiation at energy higher than in the previous experiments. Hb-Vs location in the vessel lumen was identified to explain its photosensitizer effect using human Hb immunostaining of the irradiated wattles. Laser irradiation with Hb-Vs can effectively destroy deep vessels in animal models. Hb-Vs tend to flow in the marginal zone of both small and large vessels. Increasing laser power combined with Hb-Vs injection contributed for deep vessel impairment because of the synergetic effect of both methods. Newly added Hb tended to flow near the target endothelial cells of the laser treatment. In Hb-Vs and RBC mixture, heat transfer to endothelial cells from absorbers/heater may increase. Hb-Vs function as photosensitizers to destroy deep vessels within a restricted distance that the photon can reach.

Development of a master model concept for DEMO vacuum vessel

International Nuclear Information System (INIS)

Mozzillo, Rocco; Marzullo, Domenico; Tarallo, Andrea; Bachmann, Christian; Di Gironimo, Giuseppe

2016-01-01

Highlights: • The present work concerns the development of a first master concept model for DEMO vacuum vessel. • A parametric-associative CAD master model concept of a DEMO VV sector has been developed in accordance with DEMO design guidelines. • A proper CAD design methodology has been implemented in view of the later FEM analyses based on “shell elements”. - Abstract: This paper describes the development of a master model concept of the DEMO vacuum vessel (VV) conducted within the framework of the EUROfusion Consortium. Starting from the VV space envelope defined in the DEMO baseline design 2014, the layout of the VV structure was preliminarily defined according to the design criteria provided in RCC-MRx. A surface modelling technique was adopted and efficiently linked to the finite element (FE) code to simplify future FE analyses. In view of possible changes to shape and structure during the conceptual design activities, a parametric design approach allows incorporating modifications to the model efficiently.

Development of a master model concept for DEMO vacuum vessel

Energy Technology Data Exchange (ETDEWEB)

Mozzillo, Rocco; Marzullo, Domenico; Tarallo, Andrea [CREATE, University of Naples Federico II, DII, P.le Tecchio 80, 80125, Naples (Italy); Bachmann, Christian [EUROfusion PMU, Boltzmannstraße 2, 85748 Garching (Germany); Di Gironimo, Giuseppe, E-mail: peppe.digironimo@gmail.com [CREATE, University of Naples Federico II, DII, P.le Tecchio 80, 80125, Naples (Italy)

2016-11-15

Highlights: • The present work concerns the development of a first master concept model for DEMO vacuum vessel. • A parametric-associative CAD master model concept of a DEMO VV sector has been developed in accordance with DEMO design guidelines. • A proper CAD design methodology has been implemented in view of the later FEM analyses based on “shell elements”. - Abstract: This paper describes the development of a master model concept of the DEMO vacuum vessel (VV) conducted within the framework of the EUROfusion Consortium. Starting from the VV space envelope defined in the DEMO baseline design 2014, the layout of the VV structure was preliminarily defined according to the design criteria provided in RCC-MRx. A surface modelling technique was adopted and efficiently linked to the finite element (FE) code to simplify future FE analyses. In view of possible changes to shape and structure during the conceptual design activities, a parametric design approach allows incorporating modifications to the model efficiently.

In vessel core melt progression phenomena

International Nuclear Information System (INIS)

Courtaud, M.

1993-01-01

For all light water reactor (LWR) accidents, including the so called severe accidents where core melt down can occur, it is necessary to determine the amount and characteristics of fission products released to the environment. For existing reactors this knowledge is used to evaluate the consequences and eventual emergency plans. But for future reactors safety authorities demand decrease risks and reactors designed in such a way that fission products are retained inside the containment, the last protective barrier. This requires improved understanding and knowledge of all accident sequences. In particular it is necessary to be able to describe the very complex phenomena occurring during in vessel core melt progression because they will determine the thermal and mechanical loads on the primary circuit and the timing of its rupture as well as the fission product source term. On the other hand, in case of vessel failure, knowledge of the physical and chemical state of the core melt will provide the initial conditions for analysis of ex-vessel core melt progression and phenomena threatening the containment. Finally a good understanding of in vessel phenomena will help to improve accident management procedures like Emergency Core Cooling System water injection, blowdown and flooding of the vessel well, with their possible adverse effects. Research and Development work on this subject was initiated a long time ago and is still in progress but now it must be intensified in order to meet the safety requirements of the next generation of reactors. Experiments, limited in scale, analysis of the TMI 2 accident which is a unique source of global information and engineering judgment are used to establish and assess physical models that can be implemented in computer codes for reactor accident analysis

Pressure vessel design manual

CERN Document Server

Moss, Dennis R

2013-01-01

Pressure vessels are closed containers designed to hold gases or liquids at a pressure substantially different from the ambient pressure. They have a variety of applications in industry, including in oil refineries, nuclear reactors, vehicle airbrake reservoirs, and more. The pressure differential with such vessels is dangerous, and due to the risk of accident and fatality around their use, the design, manufacture, operation and inspection of pressure vessels is regulated by engineering authorities and guided by legal codes and standards. Pressure Vessel Design Manual is a solutions-focused guide to the many problems and technical challenges involved in the design of pressure vessels to match stringent standards and codes. It brings together otherwise scattered information and explanations into one easy-to-use resource to minimize research and take readers from problem to solution in the most direct manner possible. * Covers almost all problems that a working pressure vessel designer can expect to face, with ...

Effect of air content and mass inflow on the pressure rise in a containment during blowdown

International Nuclear Information System (INIS)

Marshall, J.; Holland, P.G.

1977-01-01

Experiments were made to investigate conditions arising during blowdown of a vessel filled with saturated steam/water at 7 MPa pressure into a containment vessel. The initial air pressure in the containment vessel was varied from one atmosphere to near vacuum. The initial water content of the high pressure vessel was varied. Pressure and temperature distributions were measured during the blowdown transient and compared with calculations based on a simple lumped-parameter model. The effect of condensation heat transfer on the containment pressure is discussed and attention drawn to the inadequacy of most available data. (Author)

Study of evaluation methods for in-vessel corium retention through external vessel cooling and safety of reactor cavity

Energy Technology Data Exchange (ETDEWEB)

Huh, Hoon; Chang, Soon Heung; Kim, Soo Hyung; Kim, Kee Poong; Lee, Hyoung Wook; Jang, Kwang Keol; Jeong, Yong Hoon; Kim, Sang Jin; Lee, Seong Jin [Korea Advanced Institute of Science and Technology, Taejon (Korea, Republic of); Park, Jae Hong [Korea Institute of Nuclear Safety, Taejon (Korea, Republic of)

2001-03-15

In this work, assessment system for methodology for reactor pressure vessel integrity is developed. Assessment system is make up of severe accident assessment code which can calculate the conditions of plant and structural analysis code which can assess the integrity of reactor vessel using given plant conditions. An assessment of cavity flooding using containment spray system has been done. As a result, by the containment spray, cavity can be flooded successfully and CCI can be reduced. The technical backgrounds for external vessel cooling and corium cooling on the cavity are summarized and provided in this report.

Study of evaluation methods for in-vessel corium retention through external vessel cooling and safety of reactor cavity

International Nuclear Information System (INIS)

Huh, Hoon; Chang, Soon Heung; Kim, Soo Hyung; Kim, Kee Poong; Lee, Hyoung Wook; Jang, Kwang Keol; Jeong, Yong Hoon; Kim, Sang Jin; Lee, Seong Jin; Park, Jae Hong

2001-03-01

In this work, assessment system for methodology for reactor pressure vessel integrity is developed. Assessment system is make up of severe accident assessment code which can calculate the conditions of plant and structural analysis code which can assess the integrity of reactor vessel using given plant conditions. An assessment of cavity flooding using containment spray system has been done. As a result, by the containment spray, cavity can be flooded successfully and CCI can be reduced. The technical backgrounds for external vessel cooling and corium cooling on the cavity are summarized and provided in this report

Neutron spectrum measurement inside containment vessel at Kori nuclear power plant unit 1

Energy Technology Data Exchange (ETDEWEB)

Han, J. M.; Kim, T. W.; Kim, K. D.; Youn, C. H. [Nuclear Environment Technology Institute, Taejon (Korea, Republic of)

2003-10-01

There would be a case for the radiation worker have to work inside of the containment vessel to inspect or repair reactor facilities. In this case, the information about distribution of neutron field is needed to estimate neutron exposure dose of worker. Neutron spectra were measured by BMS(Bonner Multisphere Spectrometer) at 4 points of 6 ft and 20 ft, 2 points of 44 ft, 5 points of 70 ft in containment vessel of Kori unit 1. From the calculation, the following results were obtained. Neutron fluxes of 6 ft were between 2.623 x 10{sup 2} and 2.746 x 10{sup 4} neutron/cm{sup 2}{center_dot}sec, average neutron energies were between 9.209 x 10{sup -6} and 3.377 x 10{sup -2} MeV, equivalent doses of neutron were between 0.025 and 2.675 mSv/hr. Neutron fluxes of 20 ft were between 1.771 x 10{sup 1} and 1.682 x 10{sup 3} neutron/cm{sup 2}{center_dot}sec, average neutron energies were between 6.084 x 10{sup -6} and 2.988 x 10{sup -1} MeV, equivalent doses of neutron were between 0.004 and 0.228 mSv/hr. Neutron fluxes of 44 ft were between 3.367 x 10{sup 2} and 3.483 x 10{sup 2} neutron / cm{sup 2}{center_dot}sec, average neutron energies were between 3.962 x 10{sup -2} and 7.360 x 10{sup -2} MeV, equivalent doses of neutron were between 0.069 and 0.089 mSv/hr. Neutron fluxes of 70 ft were between 4.553 x 10{sup 3} and 1.407 x 10{sup 4} neutron/cm{sup 2}{center_dot}sec, average neutron energies were between 3.668 x 10{sup -4} and 6.764 x 10{sup -2} MeV, equivalent doses of neutron were between 0.449 and 2.660 mSv/hr.

Triple –E Vessels: Tonnage Measurement and Suez Canal Dues Assessment

Directory of Open Access Journals (Sweden)

Elsayed Hussein Galall

2015-08-01

Full Text Available Container is growing faster than GDP, Shipping lines always attempt to augment efficiencyby reducing cost and by attracting larger volumes of containers. As a result rising containerfreight rates the lines have been driven to increase economic of scale, by building mega shipsand fewer mere efficient port calls. In 2011 Maersk line ordered up to 20 new “Triple- E “Class of container vessels deliversbetween 2013- 2015. These class of mega container vessels have its way through Suez Canal,other companies CMA, CGM also ordered this type of mega container vessels, in order toreach higher profits due to the achieved economics of scale It is believed that 20000 TEUcould be the next target size. Present mega container fleet and any future feasible potential vessel capacity expansionmore than 18000 TEU put Suez Canal route in strong competitive position. MeanwhilePanama Canal will not be able to handle vessels larger than 12600 TEU even after itsexpansion in 2015.

Nonlinear analysis of end slabs in prestressed concrete reactor vessels

International Nuclear Information System (INIS)

Abdulrahman, H.O.

1978-01-01

A procedure for the nonlinear analysis of end slabs is prestressed concrete reactor vessels (PCRVs), based on the finite element method, is presented. The applicability of the procedure to the ultimate load analysis of small-scale models of the primary containment of nuclear reactors is shown. Material nonlinearity only is considered. The procedure utilizes the four-node linear quadrilateral isoparametric element with the choice of incorporating the nonconforming modes. This element is used for modeling the vessel as an axisymmetric solid. Concrete is assumed to be an isotropic material in the elastic range. The compressive stresses are judged according to a special form of the Mohr-Coulomb criterion. The nonlinear problem was solved using a generalized Newton-Raphson procedure. A detailed example problem of a pressure vessel with penetrations is presented. This is followed by a summary of the other cases studied. The solutions obtained match very closely the measured response of the test vessels under increasing internal pressure up to failure. The procedure is thus adequate for the assessment of the ultimate load behavior and failure of actual pressure vessels with a moderate demand on human and computational resources

Comparison of elastic--plastic and variable modulus-cracking constitutive models for prestressed concrete reactor vessels

International Nuclear Information System (INIS)

Anderson, C.A.; Smith, P.D.

1978-01-01

The variable modulus-cracking model is capable of predicting the behavior of reinforced concrete structures (such as the reinforced plate under transverse pressure described previously) well into the range of nonlinear behavior including the prediction of the ultimate load. For unreinforced thick-walled concrete vessels under internal pressure the use of elastic--plastic concrete models in finite element codes enhances the apparent ductility of the vessels in contrast to variable modulus-cracking models that predict nearly instantaneous rupture whenever the tensile strength at the inner wall is exceeded. For unreinforced thick-walled end slabs representative of PCRV heads, the behavior predicted by finite element codes using variable modulus-cracking models is much stiffer in the nonlinear range than that observed experimentally. Although the shear type failures and crack patterns that are observed experimentally are predicted by such concrete models, the ultimate load carrying capacity and vessel-ductility are significantly underestimated. It appears that such models do not adequately model such features as aggregate interlock that could lead to an enhanced vessel reserve strength and ductility

Parametric Study on Important Variables of Aircraft Impact to Prestressed Concrete Containment Vessels

Energy Technology Data Exchange (ETDEWEB)

Shin, Sangshup; Hahm, Daegi; Choi, Inkil [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2013-05-15

In this paper, to find the damage parameter, it is necessary to use many analysis cases and the time reduction. Thus, this paper uses a revised version of Riera's method. Using this method, the response has been found a Prestressed Concrete Containments Vessels (PCCVs) subject to impact loading, and the results of the velocity and mass of the important parameters have been analyzed. To find the response of the PCCVs subjected to aircraft impact load, it is made that a variable forcing functions depending on the velocity and fuel in the paper. The velocity variation affects more than fuel percentage, and we expect that the severe damage of the PCCVs with the same material properties is subject to aircraft impact load (more than 200m/s and 70%)

Analysis and application of prestressed concrete reactor vessels for LMFBR containment

International Nuclear Information System (INIS)

Marchertas, A.H.; Fistedis, S.H.; Bazant, Z.P.; Belytschko, T.B.

1978-01-01

An analytical model of a prestressed concrete reactor vessel (PCRV) for LMFBR and the associated finite element computer code, involving an explicit time integration procedure, is described. The model is axisymmetric and includes simulations of the tensile cracking of concrete, the reinforcement, and a prestressing capability. The tensile cracking of concrete and the steel reinforcement are both modeled as continuously distributed within the finite element. The stresses in the reinforcement and concrete are computed separately and combined to give an overall stress state of the composite material. Attention is given to the fact that cracks do not form instantaneously, but develop gradually. Thus, after crack initiation the normal stress is reduced to zero gradually as a function of time. Residual shear resistance of cracks due to aggregate interlock is also taken into account. Prestressing of the PCRV is modeled by special structural members which represent an averaged prestressing layer equivalent to an axisymmetric shell. The internal prestressing members are superimposed over the reinforced concrete body of the PCRV; they are permitted to stretch and slide in a predetermined path, simulating the actual tendons. The validity of the code is examined by comparison with experimental data. (Auth.)

Quantitative analysis of artifacts in 4D DSA: the relative contributions of beam hardening and scatter to vessel dropout behind highly attenuating structures

Science.gov (United States)

Hermus, James; Szczykutowicz, Timothy P.; Strother, Charles M.; Mistretta, Charles

2014-03-01

When performing Computed Tomographic (CT) image reconstruction on digital subtraction angiography (DSA) projections, loss of vessel contrast has been observed behind highly attenuating anatomy, such as dental implants and large contrast filled aneurysms. Because this typically occurs only in a limited range of projection angles, the observed contrast time course can potentially be altered. In this work, we have developed a model for acquiring DSA projections that models both the polychromatic nature of the x-ray spectrum and the x-ray scattering interactions to investigate this problem. In our simulation framework, scatter and beam hardening contributions to vessel dropout can be analyzed separately. We constructed digital phantoms with large clearly defined regions containing iodine contrast, bone, soft issue, titanium (dental implants) or combinations of these materials. As the regions containing the materials were large and rectangular, when the phantoms were forward projected, the projections contained uniform regions of interest (ROI) and enabled accurate vessel dropout analysis. Two phantom models were used, one to model the case of a vessel behind a large contrast filled aneurysm and the other to model a vessel behind a dental implant. Cases in which both beam hardening and scatter were turned off, only scatter was turned on, only beam hardening was turned on, and both scatter and beam hardening were turned on, were simulated for both phantom models. The analysis of this data showed that the contrast degradation is primarily due to scatter. When analyzing the aneurysm case, 90.25% of the vessel contrast was lost in the polychromatic scatter image, however only 50.5% of the vessel contrast was lost in the beam hardening only image. When analyzing the teeth case, 44.2% of the vessel contrast was lost in the polychromatic scatter image and only 26.2% of the vessel contrast was lost in the beam hardening only image.

A Novel Through Capacity Model for One-way Channel Based on Characteristics of the Vessel Traffic Flow

Directory of Open Access Journals (Sweden)

Yuanyuan Nie

2017-09-01

Full Text Available Vessel traffic flow is a key parameter for channel-through capacity and is of great significance to vessel traffic management, channel and port design and navigational risk evaluation. Based on the study of parameters of characteristics of vessel traffic flow related to channel-through capacity, this paper puts forward a brand-new mathematical model for one-way channel-through capacity in which parameters of channel length, vessel arrival rate and velocity difference in different vessels are involved and a theoretical calculating mechanism for the channel-through capacity is provided. In order to verify availability and reliability of the model, extensive simulation studies have been carried out and based on the historical AIS data, an analytical case study on the Xiazhimen Channel validating the proposed model is presented. Both simulation studies and the case study show that the proposed model is valid and all relative parameters can be readjusted and optimized to further improve the channel-through capacity. Thus, all studies demonstrate that the model is valuable for channel design and vessel management.

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

Blood Vessel Normalization in the Hamster Oral Cancer Model for Experimental Cancer Therapy Studies

Energy Technology Data Exchange (ETDEWEB)

Ana J. Molinari; Romina F. Aromando; Maria E. Itoiz; Marcela A. Garabalino; Andrea Monti Hughes; Elisa M. Heber; Emiliano C. C. Pozzi; David W. Nigg; Veronica A. Trivillin; Amanda E. Schwint

2012-07-01

Normalization of tumor blood vessels improves drug and oxygen delivery to cancer cells. The aim of this study was to develop a technique to normalize blood vessels in the hamster cheek pouch model of oral cancer. Materials and Methods: Tumor-bearing hamsters were treated with thalidomide and were compared with controls. Results: Twenty eight hours after treatment with thalidomide, the blood vessels of premalignant tissue observable in vivo became narrower and less tortuous than those of controls; Evans Blue Dye extravasation in tumor was significantly reduced (indicating a reduction in aberrant tumor vascular hyperpermeability that compromises blood flow), and tumor blood vessel morphology in histological sections, labeled for Factor VIII, revealed a significant reduction in compressive forces. These findings indicated blood vessel normalization with a window of 48 h. Conclusion: The technique developed herein has rendered the hamster oral cancer model amenable to research, with the potential benefit of vascular normalization in head and neck cancer therapy.

Containment heat removal system

International Nuclear Information System (INIS)

Wade, G.E.; Barbanti, G.; Gou, P.F.; Rao, A.S.; Hsu, L.C.

1992-01-01

This patent describes a nuclear system of a type including a containment having a nuclear reactor therein, the nuclear reactor including a pressure vessel and a core in the pressure vessel, the system. It comprises a gravity pool of coolant disposed at an elevation sufficient to permit a flow of coolant into the nuclear reactor pressure vessel against a predetermined pressure within the nuclear reactor pressure vessel; means for reducing a pressure of steam in the nuclear reactor pressure vessel to a value less than the predetermined pressure in the event of a nuclear accident, the means including a depressurization valve connected to the pressure vessel, the means further including steam heat dissipating means such dissipating means including a suppression pool; a supply of water in the suppression pool, there being a headspace in the suppression pool above the water supply; a substantial amount of air in the head space; means for feeding pressurized steam from the nuclear reactor pressure vessel to a location under a surface of the supply of water, the supply of water being effective to absorb heat sufficient to reduce steam pressure below the predetermined pressure; and a check valve for communicating the headspace with the containment, the check valve being oriented to vent air in the headspace to the containment when a pressure in the headspace exceeds a pressure in the containment by a predetermined pressure differential

Electromagnetic forces on a metallic Tokamak vacuum vessel following a disruptive instability

International Nuclear Information System (INIS)

Eckhartt, D.

1979-04-01

During a 'hard' disruptive instability of a Tokamak plasma the current-carrying plasma is lost within a very short time, typically few milliseconds. If the plasma is contained in a metallic vacuum vessel, electric currents are set up in the vessel following the disappearance of the plasma current. These vessel currents together with the magnetic fields intersecting the vessel generate electromagnetic forces which appear as mechanical loads on the vessel. In the following note it is assumed that the vacuum vessel is surrounded by an 'outer equivalent' or 'flux-conserving' shell having a characteristic time of magnetic field penetration which is long compared to the time of existence of the vessel currents. This property defines the distribution of vessel current densities (and hence the load distribution) without referring to the exact mechanism or time sequence of events by which the plasma current is lost. Numerical examples of the electromagnetic force distribution from this model refer to parameters of the JET-device with the simplifying assumption of circular cross-sections for plasma current, vacuum vessel, and outer equivalent shell. (orig.)

Design and construction of reactor containment systems of the prototype fast breeder reactor MONJU

International Nuclear Information System (INIS)

Ikeda, Makinori; Kawata, Koji; Sato, Masaki; Ito, Masashi; Hayashi, Kazutoshi; Kunishima, Shigeru.

1991-01-01

The MONJU reactor containment systems consist of a reactor containment vessel, reactor cavity walls and cell liners. The reactor containment vessel is strengthened by ring stiffeners for earthquake stresses. To verify its earthquake-resistant strength, vibration and buckling tests were carried out by using 1/19 scale models. The reactor cavity walls, which form biological shield and support the reactor vessel, are constructed of steel plate frames filled with concrete. The cell liner consists of liner plates and thermal insulation to moderate the effects of sodium spills, and forms a gastight cell to maintain a nitrogen atmosphere. (author)

Calculation method for residual stress analysis of filament-wound spherical pressure vessels

International Nuclear Information System (INIS)

Knight, C.E. Jr.

1976-01-01

Filament wound spherical pressure vessels may be produced with very high performance factors. These performance factors are a calculation of contained pressure times enclosed volume divided by structure weight. A number of parameters are important in determining the level of performance achieved. One of these is the residual stress state in the fabricated unit. A significant level of an unfavorable residual stress state could seriously impair the performance of the vessel. Residual stresses are of more concern for vessels with relatively thick walls and/or vessels constructed with the highly anisotropic graphite or aramid fibers. A method is established for measuring these stresses. A theoretical model of the composite structure is required. Data collection procedures and techniques are developed. The data are reduced by means of the model and result in the residual stress analysis. The analysis method can be used in process parameter studies to establish the best fabrication procedures

Containment Performance Evaluation of a Sodium Fire Event Due to Air Ingress into the Cover Gas Region of the Reactor Vessel in the PGSFR

International Nuclear Information System (INIS)

Ahn, Sang June; Chang, Won-Pyo; Kang, Seok Hun; Choi, Chi-Woong; Yoo, Jin; Lee, Kwi Lim; Jeong, Jae-Ho; Lee, Seung Won; Jeong, Taekyeong; Ha, Kwi-Seok

2015-01-01

Comparing with the light water reactor, sodium as a reactor coolant violently reacts with oxygen in the containment atmosphere. Due to this chemical reaction, heat generated from the combustion heat increases the temperature and pressure in the containment atmosphere. The structural integrity of the containment building which is a final radiological defense barrier is threaten. A sodium fire event in the containment due to air ingress into the cover gas region in the reactor vessel is classified as one of the design basis events in the PGSFR. This event comes from a leak or crack on the reactor upper closure header surface. It accompanys an event of the radiological fission products release to the inside the containment. In this paper, evaluation for the sodium fire and radiological influence due to air ingress into the cover gas region of the reactor vessel is described. To evaluate this event, the CONTAIN-LMR, MACCS-II and OR-IGEN-II codes are used. For the sodium pool fire event in the containment, the performance evaluation and radiological influence are carried out. In the thermal hydraulic aspects, the 1 cell containment yields the most conservative result. In this event, the maximum temperature and pressure in the containment are calculated 0.185 MPa, 280.0 .deg. C, respectively. The radiological dose at the EAB and LPZ are below the acceptance criteria specified in the 10CFR100

Containment Performance Evaluation of a Sodium Fire Event Due to Air Ingress into the Cover Gas Region of the Reactor Vessel in the PGSFR

Energy Technology Data Exchange (ETDEWEB)

Ahn, Sang June; Chang, Won-Pyo; Kang, Seok Hun; Choi, Chi-Woong; Yoo, Jin; Lee, Kwi Lim; Jeong, Jae-Ho; Lee, Seung Won; Jeong, Taekyeong; Ha, Kwi-Seok [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2015-05-15

Comparing with the light water reactor, sodium as a reactor coolant violently reacts with oxygen in the containment atmosphere. Due to this chemical reaction, heat generated from the combustion heat increases the temperature and pressure in the containment atmosphere. The structural integrity of the containment building which is a final radiological defense barrier is threaten. A sodium fire event in the containment due to air ingress into the cover gas region in the reactor vessel is classified as one of the design basis events in the PGSFR. This event comes from a leak or crack on the reactor upper closure header surface. It accompanys an event of the radiological fission products release to the inside the containment. In this paper, evaluation for the sodium fire and radiological influence due to air ingress into the cover gas region of the reactor vessel is described. To evaluate this event, the CONTAIN-LMR, MACCS-II and OR-IGEN-II codes are used. For the sodium pool fire event in the containment, the performance evaluation and radiological influence are carried out. In the thermal hydraulic aspects, the 1 cell containment yields the most conservative result. In this event, the maximum temperature and pressure in the containment are calculated 0.185 MPa, 280.0 .deg. C, respectively. The radiological dose at the EAB and LPZ are below the acceptance criteria specified in the 10CFR100.

Establishment of welding process without PWHT and preheating in SGV480 plate for nuclear reactor containment vessel

International Nuclear Information System (INIS)

Watanabe, Nozomu; Higashikubo, Tomohiro; Nagamura, Takafumi; Yoshimoto Kentaro

2000-01-01

Ordinances of Japan's Ministry of International Trade and Industry provide that welded joints more than 38 mm thick used in nuclear reactor containment vessels undergo Post Weld Heat Treatment (PWHT). PWHT is difficult to apply in the field, however. We made SGV480 plate tougher and more weldable by using a Thermo-Mechanical Control Process (TMCP) in rolling. Such plate can be used without PWHT or preheating up to 55 mm thick at lowest service temperature -19degC. (author)

Establishment of welding process without PWHT and preheating in SGV480 plate for nuclear reactor containment vessel

Energy Technology Data Exchange (ETDEWEB)

Watanabe, Nozomu; Higashikubo, Tomohiro; Nagamura, Takafumi [Mitsubishi Heavy Industries. Ltd., Kobe Shipyard and Machinery Works (Japan); Yoshimoto Kentaro [Mitsubishi Heavy Industries Ltd., Takasago, Hyogo (Japan). Takasago Research and Development Center

2000-07-01

Ordinances of Japan's Ministry of International Trade and Industry provide that welded joints more than 38 mm thick used in nuclear reactor containment vessels undergo Post Weld Heat Treatment (PWHT). PWHT is difficult to apply in the field, however. We made SGV480 plate tougher and more weldable by using a Thermo-Mechanical Control Process (TMCP) in rolling. Such plate can be used without PWHT or preheating up to 55 mm thick at lowest service temperature -19degC. (author)

Tumor Blood Vessel Dynamics

Science.gov (United States)

Munn, Lance

2009-11-01

``Normalization'' of tumor blood vessels has shown promise to improve the efficacy of chemotherapeutics. In theory, anti-angiogenic drugs targeting endothelial VEGF signaling can improve vessel network structure and function, enhancing the transport of subsequent cytotoxic drugs to cancer cells. In practice, the effects are unpredictable, with varying levels of success. The predominant effects of anti-VEGF therapies are decreased vessel leakiness (hydraulic conductivity), decreased vessel diameters and pruning of the immature vessel network. It is thought that each of these can influence perfusion of the vessel network, inducing flow in regions that were previously sluggish or stagnant. Unfortunately, when anti-VEGF therapies affect vessel structure and function, the changes are dynamic and overlapping in time, and it has been difficult to identify a consistent and predictable normalization ``window'' during which perfusion and subsequent drug delivery is optimal. This is largely due to the non-linearity in the system, and the inability to distinguish the effects of decreased vessel leakiness from those due to network structural changes in clinical trials or animal studies. We have developed a mathematical model to calculate blood flow in complex tumor networks imaged by two-photon microscopy. The model incorporates the necessary and sufficient components for addressing the problem of normalization of tumor vasculature: i) lattice-Boltzmann calculations of the full flow field within the vasculature and within the tissue, ii) diffusion and convection of soluble species such as oxygen or drugs within vessels and the tissue domain, iii) distinct and spatially-resolved vessel hydraulic conductivities and permeabilities for each species, iv) erythrocyte particles advecting in the flow and delivering oxygen with real oxygen release kinetics, v) shear stress-mediated vascular remodeling. This model, guided by multi-parameter intravital imaging of tumor vessel structure

Primo vessel inside a lymph vessel emerging from a cancer tissue.

Science.gov (United States)

Lee, Sungwoo; Ryu, Yeonhee; Cha, Jinmyung; Lee, Jin-Kyu; Soh, Kwang-Sup; Kim, Sungchul; Lim, Jaekwan

2012-10-01

Primo vessels were observed inside the lymph vessels near the caudal vena cava of a rabbit and a rat and in the thoracic lymph duct of a mouse. In the current work we found a primo vessel inside the lymph vessel that came out from the tumor tissue of a mouse. A cancer model of a nude mouse was made with human lung cancer cell line NCI-H460. We injected fluorescent nanoparticles into the xenografted tumor tissue and studied their flow in blood, lymph, and primo vessels. Fluorescent nanoparticles flowed through the blood vessels quickly in few minutes, and but slowly in the lymph vessels. The bright fluorescent signals of nanoparticles disappeared within one hour in the blood vessels but remained much longer up to several hours in the case of lymph vessels. We found an exceptional case of lymph vessels that remained bright with fluorescence up to 24 hours. After detailed examination we found that the bright fluorescence was due to a putative primo vessel inside the lymph vessel. This rare observation is consistent with Bong-Han Kim's claim on the presence of a primo vascular system in lymph vessels. It provides a significant suggestion on the cancer metastasis through primo vessels and lymph vessels. Copyright © 2012. Published by Elsevier B.V.

Development of an automated remote inspection system for the interior of the primary containment vessel of a nuclear power plant

International Nuclear Information System (INIS)

Senoo, Makoto; Yoshida, Tomiharu; Omote, Tatsuyuki; Tanaka, Keiji; Koga, Kazunori

1996-01-01

An automated remote inspection system has been developed for the interior of the primary containment vessel of a nuclear power plant. This system consists of an inspection robot and an operator's console. The inspection robot travels along a monorail provided in the interior of the primary containment vessel. The operator's console is located in the central control room of the power plant. We have made efforts to downsize the robot and automate the inspection and monitoring machinery. As for downsizing the robot, a 152 mm wide, 290 mm high cross-sectional area and 15 kg weight can be realized using commercially available small sensors and rearranging the parts in those sensors. As for automating the inspection and monitoring, several monitoring functions are developed using image processing, frequency analysis and other techniques applied to signals from sensors such as an ITV camera, an infrared camera and a microphone, which are mounted on the robot. Endurance tests show resistance of the robot to radiational and thermal conditions is adequate for actual use in actual power plants. (author)

Transient stratification modelling of a corium pool in a LWR vessel lower head

International Nuclear Information System (INIS)

Le Tellier, R.; Saas, L.; Bajard, S.

2015-01-01

Highlights: • A kinetic stratification model is proposed for the simulation of the in-vessel corium behaviour during a LWR severe accident. • The different associated “modes” of vessel failure by thermal focusing effect are highlighted and discussed. • A sensitivity study for a 1650 MWe GenIII PWR is presented with this model in order to illustrate the associated R&D issues. - Abstract: In the context of light water reactor severe accidents analysis, this paper is focused on one key parameter of in-vessel corium phenomenology: the immiscible phases stratification and its impact on the heat flux distribution at the corium pool lateral boundary with the so-called focusing effect related to a “thin” top metal phase and the potential vessel failure at that point. More particularly, based on the limited knowledge of the stratification transient phenomenon derived from the MASCA-RCW experiment, a basic model is proposed that can be used for corium in lower head sensitivity analyses. It has been implemented in the PROCOR platform developed at CEA Cadarache. A short parametric study on a simple hypothetical transient is presented in order to highlight the different focusing effect “modes” that can be encountered based on this in-vessel corium pool model. An early mode may occur during the formation of the top metal layer while two other modes may appear later during the thinning of this top metal layer because of thermochemically induced mass transfers. Some associated relevant parameters (model or scenario-dependent) and modelling issues are mentioned and illustrated with some results of a Monte-Carlo based sensitivity calculation on the transient behaviour of the corium in the lower head of a 1650 MWe GenIII PWR. Within the limiting modelling hypotheses, the thermal modelling of the steel layer for small (centimetre) heights and the mass diffusivity (limited in this case to the uranium diffusivity in the oxidic layer) are main sensitive parameters

Review of the current understanding of the potential for containment failure from in-vessel steam explosions

International Nuclear Information System (INIS)

1985-06-01

A group of experts was convened to review the current understanding of the potential for containment failure from in-vessel steam explosions during core meltdown accidents in LWRs. The Steam Explosion Review Group (SERG) was requested to provide assessments of: (1) the conditional probability of containment failure due to a steam explosion, (2) a Sandia National Laboratory (SNL) report entitled ''An Uncertainty Study of PWR Steam Explosions,'' NUREG/CR-3369, (3) a SNL proposed steam explosion research program. This report summarizes the results of the deliberations of the review group. It also presents the detailed response of each individual member to each of the issues. The consensus of the SERG is that the occurrence of a steam explosion of sufficient energetics which could lead to alpha-mode containment failure has a low probability. The SERG members disagreed with the methodology used in NUREG/CR-3369 for the purpose of establishing the uncertainty in the probability of containment failure by a steam explosion. A consensus was reached among SERG members on the need for a continuing steam explosion research program which would improve our understanding of certain aspects of steam explosion phenomenology

In-Vessel Coolability. Workshop Proceedings, in collaboration with EC-SARNET

International Nuclear Information System (INIS)

2011-01-01

Severe Accident Management Guidelines increase focus on containment integrity after some progression in the course of a severe accident. This change in priorities is made according to criteria that vary depending on reactor type and specific procedures. Once a water source has been recovered, different accident management strategies can be used: send water into the core and/or cool the reactor pressure vessel (RPV) externally. It should be noticed that, depending on the amount of water available, these strategies might conflict with other uses of water such as for instance activating spray systems in the containment or may have deleterious effects as for instance an increase in the production of hydrogen. Generally, for in-vessel reflooding, the models used for evaluation of accident management measures suffer from a lack of validation. Given this background, the objectives of the workshop were: -) to exchange information on different Severe Accident Management strategies used or contemplated for the in-vessel coolability issue; -) to review recent, ongoing and planned experimental programmes on reflooding; -) to review models used for reflooding in severe accident calculation tools, either simplified or sophisticated; -) to exchange information on the treatment of reflooding in different safety studies such as Probabilistic Safety Assessment; and -) to provide recommendations for future work, as necessary

Modeling for evaluation of debris coolability in lower plenum of reactor pressure vessel

International Nuclear Information System (INIS)

Maruyama, Yu; Moriyama, Kiyofumi; Nakamura, Hideo; Hirano, Masashi

2003-01-01

Effectiveness of debris cooling by water that fills a gap between the debris and the lower head wall was estimated through steady calculations in reactor scale. In those calculations, the maximum coolable debris depth was assessed as a function of gap width with combination of correlations for critical heat flux and turbulent natural convection of a volumetrically heated pool. The results indicated that the gap with a width of 1 to 2 mm was capable of cooling the debris under the conditions of the TMI-2 accident, and that a significantly larger gap width was needed to retain a larger amount of debris within the lower plenum. Transient models on gap growth and water penetration into the gap were developed and incorporated into CAMP code along with turbulent natural convection model developed by Yin, Nagano and Tsuji, categorized in low Reynolds number type two-equation model. The validation of the turbulent model was made with the UCLA experiment on natural convection of a volumetrically heated pool. It was confirmed that CAMP code predicted well the distribution of local heat transfer coefficients along the vessel inner surface. The gap cooling model was validated by analyzing the in-vessel debris coolability experiments at JAERI, where molten Al 2 O 3 was poured into a water-filled hemispherical vessel. The temperature history measured on the vessel outer surface was satisfactorily reproduced by CAMP code. (author)

Study of the concrete tensile creep: application for the containment vessel of the nuclear power plants (PWR)

International Nuclear Information System (INIS)

Reviron, Nanthilde

2009-01-01

The aim of this work is to study experimentally and to conduct numerical simulations on the creep of concrete subjected to tensile stresses. The main purpose is to predict the behaviour of containment vessels of nuclear power plants (PWR) in the case of decennial test or accident. In order to satisfy to these industrial needs, it is necessary to characterize the behaviour of concrete under uniaxial tension. Thus, an important experimental study of tensile creep in concrete has been performed for different loading levels (50%, 70% and 90% of the tensile strength). In these tests, load was kept constant during 3 days. Several tests were performed: measurements of elastic properties and strength (in tension and in compression), monitoring of drying, shrinkage, basic creep and drying creep strains. Moreover, compressive creep tests were also performed and showed a difference with tensile creep. Furthermore, decrease of tensile strength and failure under tensile creep for large loading levels were observed. A numerical model has been proposed and developed in Cast3m finite element code. (author)

A Fovea Localization Scheme Using Vessel Origin-Based Parabolic Model

Directory of Open Access Journals (Sweden)

Chun-Yuan Yu

2014-09-01

Full Text Available At the center of the macula, fovea plays an important role in computer-aided diagnosis. To locate the fovea, this paper proposes a vessel origin (VO-based parabolic model, which takes the VO as the vertex of the parabola-like vasculature. Image processing steps are applied to accurately locate the fovea on retinal images. Firstly, morphological gradient and the circular Hough transform are used to find the optic disc. The structure of the vessel is then segmented with the line detector. Based on the characteristics of the VO, four features of VO are extracted, following the Bayesian classification procedure. Once the VO is identified, the VO-based parabolic model will locate the fovea. To find the fittest parabola and the symmetry axis of the retinal vessel, an Shift and Rotation (SR-Hough transform that combines the Hough transform with the shift and rotation of coordinates is presented. Two public databases of retinal images, DRIVE and STARE, are used to evaluate the proposed method. The experiment results show that the average Euclidean distances between the located fovea and the fovea marked by experts in two databases are 9.8 pixels and 30.7 pixels, respectively. The results are stronger than other methods and thus provide a better macular detection for further disease discovery.

Validation of ASTEC V2 models for the behaviour of corium in the vessel lower head

International Nuclear Information System (INIS)

Carénini, L.; Fleurot, J.; Fichot, F.

2014-01-01

The paper is devoted to the presentation of validation cases carried out for the models describing the corium behaviour in the “lower plenum” of the reactor vessel implemented in the V2.0 version of the ASTEC integral code, jointly developed by IRSN (France) and GRS (Germany). In the ASTEC architecture, these models are grouped within the single ICARE module and they are all activated in typical accident scenarios. Therefore, it is important to check the validity of each individual model, as long as experiments are available for which a single physical process is involved. The results of ASTEC applications against the following experiments are presented: FARO (corium jet fragmentation), LIVE (heat transfer between a molten pool and the vessel), MASCA (separation and stratification of corium non miscible phases) and OLHF (mechanical failure of the vessel). Compared to the previous ASTEC V1.3 version, the validation matrix is extended. This work allows determining recommended values for some model parameters (e.g. debris particle size in the fragmentation model and criterion for debris bed liquefaction). Almost all the processes governing the corium behaviour, its thermal interaction with the vessel wall and the vessel failure are modelled in ASTEC and these models have been assessed individually with satisfactory results. The main uncertainties appear to be related to the calculation of transient evolutions

Nuclear reactor vessel decontamination systems

International Nuclear Information System (INIS)

McGuire, P. J.

1985-01-01

There is disclosed in the present application, a decontamination system for reactor vessels. The system is operatable without entry by personnel into the contaminated vessel before the decontamination operation is carried out and comprises an assembly which is introduced into the vertical cylindrical vessel of the typical boiling water reactor through the open top. The assembly includes a circular track which is centered by guideways permanently installed in the reactor vessel and the track guides opposed pairs of nozzles through which water under very high pressure is directed at the wall for progressively cutting and sweeping a tenacious radioactive coating as the nozzles are driven around the track in close proximity to the vessel wall. The whole assembly is hoisted to a level above the top of the vessel by a crane, outboard slides on the assembly brought into engagement with the permanent guideways and the assembly progressively lowered in the vessel as the decontamination operation progresses. The assembly also includes a low pressure nozzle which forms a spray umbrella above the high pressure nozzles to contain radioactive particles dislodged during the decontamination

Development of containers sealing system like part of surveillance program of the vessel in nuclear power plants; Desarrollo del sistema de sellado de contenedores como parte del programa de vigilancia de la vasija en nucleoelectricas

Energy Technology Data Exchange (ETDEWEB)

Romero C, J.; Hernandez C, R.; Fernandez T, F.; Rocamontes A, M.; Perez R, N. [ININ, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico)], e-mail: jesus.romero@inin.gob.mx

2009-10-15

The owners of nuclear power plants should be demonstrate that the embrittlement effects by neutronic radiation do not commit the structural integrity from the pressure vessel of nuclear reactors, during conditions of routine operation and below postulate accident. For this reason, there are surveillance programs of vessels of nuclear power plants, in which are present surveillance capsules. A surveillance capsule is compound by the support, six containers for test tubes and dosimeters. The containers for test tubes are of two types: rectangular container for test tubes, Charpy V and Cylindrical Container for tension test tubes. These test tubes are subject to a same or bigger neutronic flow to that of vessel, being representative of vessel mechanical conditions. The test tubes are rehearsed to watch over the increase of embrittlement that presents the vessel. This work describes the development of welding system to seal the containers for test tubes, these should be filled with helium of ultra high purity, to a pressure of an atmosphere. In this system the welding process Gas Tungsten Arc Welding is used, a hermetic camera that allows to place the containers with three grades of freedom, a vacuum subsystem and pressure, high technology equipment's like: power source with integrated computer, arc starter of high frequency, helium flow controller, among others. Finally, the advances in the inspection system for the qualification of sealing system are mentioned, system that should measure the internal pressure of containers and the helium purity inside these. (Author)

Mathematical modelling for trajectories of magnetic nanoparticles in a blood vessel under magnetic field

International Nuclear Information System (INIS)

Sharma, Shashi; Katiyar, V.K.; Singh, Uaday

2015-01-01

A mathematical model is developed to describe the trajectories of a cluster of magnetic nanoparticles in a blood vessel for the application of magnetic drug targeting (MDT). The magnetic nanoparticles are injected into a blood vessel upstream from a malignant tissue and are captured at the tumour site with help of an applied magnetic field. The applied field is produced by a rare earth cylindrical magnet positioned outside the body. All forces expected to significantly affect the transport of nanoparticles were incorporated, including magnetization force, drag force and buoyancy force. The results show that particles are slow down and captured under the influence of magnetic force, which is responsible to attract the magnetic particles towards the magnet. It is optimized that all particles are captured either before or at the centre of the magnet (z≤0) when blood vessel is very close proximity to the magnet (d=2.5 cm). However, as the distance between blood vessel and magnet (d) increases (above 4.5 cm), the magnetic nanoparticles particles become free and they flow away down the blood vessel. Further, the present model results are validated by the simulations performed using the finite element based COMSOL software. - Highlights: • A mathematical model is developed to describe the trajectories of magnetic nanoparticles. • The dominant magnetic, drag and buoyancy forces are considered. • All particles are captured when distance between blood vessel and magnet (d) is up to 4.5 cm. • Further increase in d value (above 4.5 cm) results the free movement of magnetic particles

In-Vessel Retention Modeling Capabilities of SCDAP/RELAP5-3DC

International Nuclear Information System (INIS)

Knudson, D.L.; Rempe, J.L.

2002-01-01

Molten core materials may relocate to the lower head of a reactor vessel in the latter stages of a severe accident. Under such circumstances, in-vessel retention (IVR) of the molten materials is a vital step in mitigating potential severe accident consequences. Whether IVR occurs depends on the interactions of a number of complex processes including heat transfer inside the accumulated molten pool, heat transfer from the molten pool to the reactor vessel (and to overlying fluids), and heat transfer from exterior vessel surfaces. SCDAP/RELAP5-3D C has been developed at the Idaho National Engineering and Environmental Laboratory to facilitate simulation of the processes affecting the potential for IVR, as well as processes involved in a wide variety of other reactor transients. In this paper, current capabilities of SCDAP/RELAP5-3D C relative to IVR modeling are described and results from typical applications are provided. In addition, anticipated developments to enhance IVR simulation with SCDAP/RELAP5-3D C are outlined. (authors)

Experiments to investigate direct containment heating phenomena with scaled models of the Zion Nuclear Power Plant in the Surtsey Test Facility

International Nuclear Information System (INIS)

Allen, M.D.; Pilch, M.M.; Blanchat, T.K.; Griffith, R.O.; Nichols, R.T.

1994-05-01

The Surtsey Facility at Sandia National Laboratories (SNL) is used to perform scaled experiments that simulate hypothetical high-pressure melt ejection (HPME) accidents in a nuclear power plant (NPP). These experiments are designed to investigate the effect of specific phenomena associated with direct containment heating (DCH) on the containment load, such as the effect of physical scale, prototypic subcompartment structures, water in the cavity, and hydrogen generation and combustion. In the Integral Effects Test (IET) series, 1:10 linear scale models of the Zion NPP structures were constructed in the Surtsey vessel. The RPV was modeled with a steel pressure vessel that had a hemispherical bottom head, which had a 4-cm hole in the bottom head that simulated the final ablated hole that would be formed by ejection of an instrument guide tube in a severe NPP accident. Iron/alumina/chromium thermite was used to simulate molten corium that would accumulate on the bottom head of an actual RPV. The chemically reactive melt simulant was ejected by high-pressure steam from the RPV model into the scaled reactor cavity. Debris was then entrained through the instrument tunnel into the subcompartment structures and the upper dome of the simulated reactor containment building. The results of the IET experiments are given in this report

Experiments to investigate direct containment heating phenomena with scaled models of the Zion Nuclear Power Plant in the Surtsey Test Facility

Energy Technology Data Exchange (ETDEWEB)

Allen, M.D.; Pilch, M.M.; Blanchat, T.K.; Griffith, R.O. [Sandia National Labs., Albuquerque, NM (United States); Nichols, R.T. [Ktech Corp., Albuquerque, NM (United States)

1994-05-01

The Surtsey Facility at Sandia National Laboratories (SNL) is used to perform scaled experiments that simulate hypothetical high-pressure melt ejection (HPME) accidents in a nuclear power plant (NPP). These experiments are designed to investigate the effect of specific phenomena associated with direct containment heating (DCH) on the containment load, such as the effect of physical scale, prototypic subcompartment structures, water in the cavity, and hydrogen generation and combustion. In the Integral Effects Test (IET) series, 1:10 linear scale models of the Zion NPP structures were constructed in the Surtsey vessel. The RPV was modeled with a steel pressure vessel that had a hemispherical bottom head, which had a 4-cm hole in the bottom head that simulated the final ablated hole that would be formed by ejection of an instrument guide tube in a severe NPP accident. Iron/alumina/chromium thermite was used to simulate molten corium that would accumulate on the bottom head of an actual RPV. The chemically reactive melt simulant was ejected by high-pressure steam from the RPV model into the scaled reactor cavity. Debris was then entrained through the instrument tunnel into the subcompartment structures and the upper dome of the simulated reactor containment building. The results of the IET experiments are given in this report.

Novel method for edge detection of retinal vessels based on the model of the retinal vascular network and mathematical morphology

Science.gov (United States)

Xu, Lei; Zheng, Xiaoxiang; Zhang, Hengyi; Yu, Yajun

1998-09-01

Accurate edge detection of retinal vessels is a prerequisite for quantitative analysis of subtle morphological changes of retinal vessels under different pathological conditions. A novel method for edge detection of retinal vessels is presented in this paper. Methods: (1) Wavelet-based image preprocessing. (2) The signed edge detection algorithm and mathematical morphological operation are applied to get the approximate regions that contain retinal vessels. (3) By convolving the preprocessed image with a LoG operator only on the detected approximate regions of retinal vessels, followed by edges refining, clear edge maps of the retinal vessels are fast obtained. Results: A detailed performance evaluation together with the existing techniques is given to demonstrate the strong features of our method. Conclusions: True edge locations of retinal vessels can be fast detected with continuous structures of retinal vessels, less non- vessel segments left and insensitivity to noise. The method is also suitable for other application fields such as road edge detection.

46 CFR 115.812 - Pressure vessels and boilers.

Science.gov (United States)

2010-10-01

... 46 Shipping 4 2010-10-01 2010-10-01 false Pressure vessels and boilers. 115.812 Section 115.812... CERTIFICATION Material Inspections § 115.812 Pressure vessels and boilers. (a) Pressure vessels must be tested... testing requirements for boilers are contained in § 61.05 in subchapter F of this chapter. [CGD 85-080, 61...

Perturbation of baseline thermal stress in the Mound 9516 Shipping Package primary containment vessel

International Nuclear Information System (INIS)

Sansalone, K.H.F.

1995-01-01

Full-capacity loading of heat sources into the Mound 9516 Shipping Package primary containment vessel (PCV) results in temperature gradients which are symmetric, due to the axisymmetry of the package design. Concern over the change in thermal gradients (and therefore, stress) in the PCV due to sub-capacity loading led to the analytical examination of this phenomenon. The PCVs are cylindrical in shape and are loaded into the package such that they and all containment components are concentrically arranged along a common longitudinal axis. If the design full-capacity loading of the PCVs in this package assumes the axisymmetric (or more precisely, cyclicly symmetric) arrangement of its heat-producing contents, then sub-capacity loading implies that in many cases, the load arrangement could be asymmetric with respect to the longitudinal axis. It is then feasible that the departure from heat load axisymmetry could perturb the nominal thermal gradients so that thermally-induced stress within the PCV might increase to levels deemed unacceptable. This study applies Finite Element analysis (FEA) to the problem and demonstrates that no such unacceptable thermal stress increase occurs in the PCV material due to the asymmetric arrangement of contents. copyright 1995 American Institute of Physics

Modelling of water sump evaporation in a CFD code for nuclear containment studies

Energy Technology Data Exchange (ETDEWEB)

Malet, J., E-mail: jeanne.malet@irsn.f [Institute for Radioprotection and Nuclear Safety, DSU/SERAC/LEMAC, BP68 - 91192 Gif-sur-Yvette cedex (France); Bessiron, M., E-mail: matthieu.bessiron@irsn.f [Institute for Radioprotection and Nuclear Safety, DSU/SERAC/LEMAC, BP68 - 91192 Gif-sur-Yvette cedex (France); Perrotin, C., E-mail: christophe.perrotin@irsn.f [Institute for Radioprotection and Nuclear Safety, DSU/SERAC/LEMAC, BP68 - 91192 Gif-sur-Yvette cedex (France)

2011-05-15

Highlights: We model sump evaporation in the reactor containment for CFD codes. The sump is modelled by an interface temperature and an evaporation mass flow-rate. These two variables are modelled using energy and mass balance. Results are compared with specific experiments in a 7 m3 vessel (Tonus Qualification ANalytique, TOSQAN). A good agreement is observed, for pressure, temperatures, mass flow-rates. - Abstract: During the course of a hypothetical severe accident in a pressurized water reactor (PWR), water can be collected in the sump containment through steam condensation on walls and spray systems activation. This water is generally under evaporation conditions. The objective of this paper is twofold: to present a sump model developed using external user-defined functions for the TONUS-CFD code and to perform a first detailed comparison of the model results with experimental data. The sump model proposed here is based on energy and mass balance and leads to a good agreement between the numerical and the experimental results. Such a model can be rather easily added to any CFD code for which boundary conditions, such as injection temperature and mass flow-rate, can be modified by external user-defined functions, depending on the atmosphere conditions.

Reactor pressure vessel design

International Nuclear Information System (INIS)

Foehl, J.

1998-01-01

As a result of the popularity of the Agencies report 'Neutron Irradiation Embrittlement of Reactor Pressure Vessel Steels' of 1975, it was decided that another report on this broad subject would be of use. In this report, background and contemporary views on specially identified areas of the subject are considered as self-contained chapters, written by experts. In chapter 2, the general principles of reactor pressure vessel design are elaborated. Crack and fracture initiation and propagation are treated in some detail

Applied model of through-wall crack of coolant vessels of WWER-type reactors

International Nuclear Information System (INIS)

Petrosyan, V.; Hovakimyan, T.; Vardanyan, M.; Khachatryan, A.; Minasyan, K.

2010-01-01

We propose an applied-model of Through-Wall Crack (TWC) for WWER-type units primary vessels. The model allows to simulate the main morphological parameters of real TWC, i.e. length, area of inlet and outlet openings, channel depth and small and large size unevenness of the crack surface. The model can be used for developing and improving the coolant-leak detectors for the primary circuit vessels of WWER-units. Also, it can be used for research of the coolant two-phase leakage phenomenon through narrow cracks/channels and thermo-physical processes in heat-insulation layer of the Main Coolant Piping (MCP) during the leak

An agent-based model of the response to angioplasty and bare-metal stent deployment in an atherosclerotic blood vessel.

Directory of Open Access Journals (Sweden)

Antonia E Curtin

Full Text Available PURPOSE: While animal models are widely used to investigate the development of restenosis in blood vessels following an intervention, computational models offer another means for investigating this phenomenon. A computational model of the response of a treated vessel would allow investigators to assess the effects of altering certain vessel- and stent-related variables. The authors aimed to develop a novel computational model of restenosis development following an angioplasty and bare-metal stent implantation in an atherosclerotic vessel using agent-based modeling techniques. The presented model is intended to demonstrate the body's response to the intervention and to explore how different vessel geometries or stent arrangements may affect restenosis development. METHODS: The model was created on a two-dimensional grid space. It utilizes the post-procedural vessel lumen diameter and stent information as its input parameters. The simulation starting point of the model is an atherosclerotic vessel after an angioplasty and stent implantation procedure. The model subsequently generates the final lumen diameter, percent change in lumen cross-sectional area, time to lumen diameter stabilization, and local concentrations of inflammatory cytokines upon simulation completion. Simulation results were directly compared with the results from serial imaging studies and cytokine levels studies in atherosclerotic patients from the relevant literature. RESULTS: The final lumen diameter results were all within one standard deviation of the mean lumen diameters reported in the comparison studies. The overlapping-stent simulations yielded results that matched published trends. The cytokine levels remained within the range of physiological levels throughout the simulations. CONCLUSION: We developed a novel computational model that successfully simulated the development of restenosis in a blood vessel following an angioplasty and bare-metal stent deployment based on

Fukushima Daiichi Unit 1 Ex-Vessel Prediction: Core Concrete Interaction

International Nuclear Information System (INIS)

Robb, Kevin R; Farmer, Mitchell; Francis, Matthew W

2015-01-01

Lower head failure and corium concrete interaction were predicted to occur at Fukushima Daiichi Unit 1 (1F1) by several different system-level code analyses, including MELCOR v2.1 and MAAP5. Although these codes capture a wide range of accident phenomena, they do not contain detailed models for ex-vessel core melt behavior. However, specialized codes exist for analysis of ex-vessel melt spreading (e.g., MELTSPREAD) and long-term debris coolability (e.g., CORQUENCH). On this basis, an analysis was carried out to further evaluate ex-vessel behavior for 1F1 using MELTSPREAD and CORQUENCH. Best-estimate melt pour conditions predicted by MELCOR v2.1 and MAAP5 were used as input. MELTSPREAD was then used to predict the spatially dependent melt conditions and extent of spreading during relocation from the vessel. The results of the MELTSPREAD analysis are reported in a companion paper. This information was used as input for the long-term debris coolability analysis with CORQUENCH.

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

Overview of containment integrity test at NUPEC

International Nuclear Information System (INIS)

Takumi, K.; Yamada, T.

2004-01-01

NUPEC has started NUPEC Containment Integrity project entitled 'Proving Test on the Reliability for Reactor Containment Vessel' since June 1987. This is the project for the term of twelve years sponsored by MITI (Ministry of International Trade and Industry, Japanese Government). The test items are (1) Hydrogen mixing and distribution test, (2) Hydrogen Burning Test, (3) Iodine trapping characteristics test, and (4) Structural behavior test. Based on the test results, computer codes are verified and as the results of analysis and evaluation by the computer codes, containment integrity is to be confirmed. This paper indicates the results of hydrogen mixing and distribution test and hydrogen burning test. The NUPEC tests conducted so far suggest that hydrogen will be well mixed in the model containment vessel and the prediction by the computer code is in excellent agreement with the data. The NUPEC hydrogen burning test data is in good agreement with the FITS data at SNL that were obtained at the lower hydrogen concentration condition. (author)

Comparison of transient PCRV model test results with analysis

International Nuclear Information System (INIS)

Marchertas, A.H.; Belytschko, T.B.

1979-01-01

Comparisons are made of transient data derived from simple models of a reactor containment vessel with analytical solutions. This effort is a part of the ongoing process of development and testing of the DYNAPCON computer code. The test results used in these comparisons were obtained from scaled models of the British sodium cooled fast breeder program. The test structure is a scaled model of a cylindrically shaped reactor containment vessel made of concrete. This concrete vessel is prestressed axially by holddown bolts spanning the top and bottom slabs along the cylindrical walls, and is also prestressed circumferentially by a number of cables wrapped around the vessel. For test purposes this containment vessel is partially filled with water, which comes in direct contact with the vessel walls. The explosive charge is immersed in the pool of water and is centrally suspended from the top of the vessel. The tests are very similar to the series of tests made for the COVA experimental program, but the vessel here is the prestressed concrete container. (orig.)

Simulation-Based Optimization for Storage Allocation Problem of Outbound Containers in Automated Container Terminals

Directory of Open Access Journals (Sweden)

Ning Zhao

2015-01-01

Full Text Available Storage allocation of outbound containers is a key factor of the performance of container handling system in automated container terminals. Improper storage plans of outbound containers make QC waiting inevitable; hence, the vessel handling time will be lengthened. A simulation-based optimization method is proposed in this paper for the storage allocation problem of outbound containers in automated container terminals (SAPOBA. A simulation model is built up by Timed-Colored-Petri-Net (TCPN, used to evaluate the QC waiting time of storage plans. Two optimization approaches, based on Particle Swarm Optimization (PSO and Genetic Algorithm (GA, are proposed to form the complete simulation-based optimization method. Effectiveness of this method is verified by experiment, as the comparison of the two optimization approaches.

Fast Generation of Container Vessel Stowage Plans:Using mixed integer programming for optimal master planning and constraint based local search for slot planning

OpenAIRE

Pacino, Dario

2012-01-01

Containerization has changed the way the world perceives shipping. It is now possible to establish complex international supply chains that have minimized shipping costs. Over the past two decades, the demand for cost efficient containerized transportation has seen a continuous increase. In order to answer to this demand, shipping companies have deployed bigger container vessels, that nowadays can transport up to 18,000 containers and are wider than the extended Panama Canal. Like busses, con...

Flexible Composite-Material Pressure Vessel

Science.gov (United States)

Brown, Glen; Haggard, Roy; Harris, Paul A.

2003-01-01

A proposed lightweight pressure vessel would be made of a composite of high-tenacity continuous fibers and a flexible matrix material. The flexibility of this pressure vessel would render it (1) compactly stowable for transport and (2) more able to withstand impacts, relative to lightweight pressure vessels made of rigid composite materials. The vessel would be designed as a structural shell wherein the fibers would be predominantly bias-oriented, the orientations being optimized to make the fibers bear the tensile loads in the structure. Such efficient use of tension-bearing fibers would minimize or eliminate the need for stitching and fill (weft) fibers for strength. The vessel could be fabricated by techniques adapted from filament winding of prior composite-material vessels, perhaps in conjunction with the use of dry film adhesives. In addition to the high-bias main-body substructure described above, the vessel would include a low-bias end substructure to complete coverage and react peak loads. Axial elements would be overlaid to contain damage and to control fiber orientation around side openings. Fiber ring structures would be used as interfaces for connection to ancillary hardware.

Quantification of the ex-vessel severe accident risks for the Swedish boiling water reactors. A scoping study performed for the APRI project

International Nuclear Information System (INIS)

Okkonen, T.; Dinh, T.N.; Bui, V.A.; Sehgal, B.R.

1995-07-01

Results of a scoping study to quantify the ex-vessel severe accident risks for the Swedish BWRs are reported. The study considers that a pool of water is established in the containment prior to vessel failure, as prescribed by the accident management scheme for the newer Swedish BWRs. The integrated methodology developed and employed combines probabilistic and deterministic treatment of the various melt-structure-water interaction processes occurring in sequence. The potential steam explosion, and the melt attack on the containment basemat, are treated with enveloping analyses. Uncertain parameters in the models and the initial conditions are treated with Monte Carlo simulations. Independent models are developed for melt coolability and possible attack on the concrete basemat. It is found that, with current models, the melt discharge scenarios, in which a large amount of accumulated melt may be released from the vessel, could subject the containment to large steam explosion loads. However, the uncertainties are so large that no definite conclusion can be drawn. The assessment of ex-vessel core debris coolability is disturbed by similar phenomenological uncertainties. Presently, coolability of the core debris can not be demonstrated. 133 refs

Quantification of the ex-vessel severe accident risks for the Swedish boiling water reactors. A scoping study performed for the APRI project

Energy Technology Data Exchange (ETDEWEB)

Okkonen, T; Dinh, T N; Bui, V A; Sehgal, B R [Royal Inst. of Tech., Stockholm (Sweden). Dept. of Energy Systems Technology

1995-07-01

Results of a scoping study to quantify the ex-vessel severe accident risks for the Swedish BWRs are reported. The study considers that a pool of water is established in the containment prior to vessel failure, as prescribed by the accident management scheme for the newer Swedish BWRs. The integrated methodology developed and employed combines probabilistic and deterministic treatment of the various melt-structure-water interaction processes occurring in sequence. The potential steam explosion, and the melt attack on the containment basemat, are treated with enveloping analyses. Uncertain parameters in the models and the initial conditions are treated with Monte Carlo simulations. Independent models are developed for melt coolability and possible attack on the concrete basemat. It is found that, with current models, the melt discharge scenarios, in which a large amount of accumulated melt may be released from the vessel, could subject the containment to large steam explosion loads. However, the uncertainties are so large that no definite conclusion can be drawn. The assessment of ex-vessel core debris coolability is disturbed by similar phenomenological uncertainties. Presently, coolability of the core debris can not be demonstrated. 133 refs.

Processing vessel for high level radioactive wastes

International Nuclear Information System (INIS)

Maekawa, Hiromichi

1998-01-01

Upon transferring an overpack having canisters containing high level radioactive wastes sealed therein and burying it into an underground processing hole, an outer shell vessel comprising a steel plate to be fit and contained in the processing hole is formed. A bury-back layer made of dug earth and sand which had been discharged upon forming the processing hole is formed on the inner circumferential wall of the outer shell vessel. A buffer layer having a predetermined thickness is formed on the inner side of the bury-back layer, and the overpack is contained in the hollow portion surrounded by the layer. The opened upper portion of the hollow portion is covered with the buffer layer and the bury-back layer. Since the processing vessel having a shielding performance previously formed on the ground, the state of packing can be observed. In addition, since an operator can directly operates upon transportation and burying of the high level radioactive wastes, remote control is no more necessary. (T.M.)

Lower Length Scale Model Development for Embrittlement of Reactor Presure Vessel Steel

Energy Technology Data Exchange (ETDEWEB)

Zhang, Yongfeng [Idaho National Lab. (INL), Idaho Falls, ID (United States); Schwen, Daniel [Idaho National Lab. (INL), Idaho Falls, ID (United States); Chakraborty, Pritam [Idaho National Lab. (INL), Idaho Falls, ID (United States); Bai, Xianming [Idaho National Lab. (INL), Idaho Falls, ID (United States)

2016-09-01

This report summarizes the lower-length-scale effort during FY 2016 in developing mesoscale capabilities for microstructure evolution, plasticity and fracture in reactor pressure vessel steels. During operation, reactor pressure vessels are subject to hardening and embrittlement caused by irradiation induced defect accumulation and irradiation enhanced solute precipitation. Both defect production and solute precipitation start from the atomic scale, and manifest their eventual effects as degradation in engineering scale properties. To predict the property degradation, multiscale modeling and simulation are needed to deal with the microstructure evolution, and to link the microstructure feature to material properties. In this report, the development of mesoscale capabilities for defect accumulation and solute precipitation are summarized. A crystal plasticity model to capture defect-dislocation interaction and a damage model for cleavage micro-crack propagation is also provided.

Structural Integrity Evaluation of Containment Vessel under Severe Accident for PGSFR

International Nuclear Information System (INIS)

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

2016-01-01

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

Cold source vessel development for the advanced neutron source

Energy Technology Data Exchange (ETDEWEB)

Williams, P.T.; Lucas, A.T. [Oak Ridge National Lab., TN (United States)

1995-09-01

The Advanced Neutron Source (ANS), in its conceptual design phase at Oak Ridge National Laboratory (ORNL), will be a user-oriented neutron research facility that will produce the most intense flux of neutrons in the world. Among its many scientific applications, the productions of cold neutrons is a significant research mission for the ANS. The cold neutrons come from two independent cold sources positioned near the reactor core. Contained by an aluminum alloy vessel, each cold source is a 410 mm diameter sphere of liquid deuterium that functions both as a neutron moderator and a cryogenic coolant. With nuclear heating of the containment vessel and internal baffling, steady-state operation requires close control of the liquid deuterium flow near the vessel`s inner surface. Preliminary thermal-hydraulic analyses supporting the cold source design are being performed with multi-dimensional computational fluid dynamics simulations of the liquid deuterium flow and heat transfer. This paper presents the starting phase of a challenging program and describes the cold source conceptual design, the thermal-hydraulic feasibility studies of the containment vessel, and the future computational and experimental studies that will be used to verify the final design.

Nuclear reactor installation with outer shell enclosing a primary pressure vessel

International Nuclear Information System (INIS)

1975-01-01

The high temperature nuclear reactor installation described includes a fluid cooled nuclear heat source, a primary pressure vessel containing the heat source, an outer shell enclosing the primary pressure vessel and acting as a secondary means of containment for this vessel against outside projectiles. Multiple auxiliary equipment points are arranged outside the outer shell which comprises a part of a lower wall around the primary pressure vessel, an annular part integrated in the lower wall and extending outwards as from this wall and an upper part integrated in the annular part and extending above this annular part and above the primary pressure vessel. The annular part and the primary pressure vessel are formed with vertical penetrations which can be closed communicating respectively with the auxiliary equipment points and with inside the pressure vessel whilst handling gear is provided in the upper part for vertically raising reactor components through these penetrations and for transporting them over the annular part and over the primary pressure vessel [fr

Evaluation for In-Vessel Retention Capabilities with In-Vessel Injection and External Reactor Vessel Cooling

International Nuclear Information System (INIS)

Lee, Jeong Seong; Ryu, In Chul; Moon, Young Tae

2016-01-01

If the accident has not progressed to the point of substantial changes in the core geometry, establishing adequate cooling is as straightforward as re-establishing flow through the reactor core. However, if the accident has progressed to the point where the core geometry is substantially altered as a result of material melting and relocation, as was the case in the TMI-2 accident, the means of cooling the debris are not as straightforward. From this time on, the reactor core was either completely or nearly covered by water, with high pressure injection flow initiated shortly after three hours into the accident. However, the core debris was not coolable in this configuration and a substantial quantity of molten core material drained into the bypass region, with approximately twenty metric tons of molten debris draining into the reactor pressure vessel (RPV) lower head. Hence, the core configuration developed at approximately three hours into the accident was not coolable, even submerged in water. The purpose of this paper is to evaluate in-vessel retention capabilities with in-vessel injection (IVI) and external reactor vessel cooling (ERVC) available in a reactor application by using the integrated severe accident analysis code. The MAAP5 models were improved to facilitate evaluation of the in-vessel retention capability of APR1400. In-vessel retention capabilities have been analyzed for the APR1400 using the MAAP5.03 code. The results show that in-vessel retention is feasible when in-vessel injection is initiated within a relatively short time frame under the simulation condition used in the present study

Evaluation for In-Vessel Retention Capabilities with In-Vessel Injection and External Reactor Vessel Cooling

Energy Technology Data Exchange (ETDEWEB)

Lee, Jeong Seong; Ryu, In Chul; Moon, Young Tae [KEPCO Engineering and Construction Co. Ltd., Deajeon (Korea, Republic of)

2016-10-15

If the accident has not progressed to the point of substantial changes in the core geometry, establishing adequate cooling is as straightforward as re-establishing flow through the reactor core. However, if the accident has progressed to the point where the core geometry is substantially altered as a result of material melting and relocation, as was the case in the TMI-2 accident, the means of cooling the debris are not as straightforward. From this time on, the reactor core was either completely or nearly covered by water, with high pressure injection flow initiated shortly after three hours into the accident. However, the core debris was not coolable in this configuration and a substantial quantity of molten core material drained into the bypass region, with approximately twenty metric tons of molten debris draining into the reactor pressure vessel (RPV) lower head. Hence, the core configuration developed at approximately three hours into the accident was not coolable, even submerged in water. The purpose of this paper is to evaluate in-vessel retention capabilities with in-vessel injection (IVI) and external reactor vessel cooling (ERVC) available in a reactor application by using the integrated severe accident analysis code. The MAAP5 models were improved to facilitate evaluation of the in-vessel retention capability of APR1400. In-vessel retention capabilities have been analyzed for the APR1400 using the MAAP5.03 code. The results show that in-vessel retention is feasible when in-vessel injection is initiated within a relatively short time frame under the simulation condition used in the present study.

Early Detection of Parametric Roll Resonance on Container Ships

DEFF Research Database (Denmark)

Galeazzi, Roberto; Blanke, Mogens; Poulsen, Niels Kjølstad

2013-01-01

Parametric roll resonance on ships is a nonlinear phenomenon where waves encountered at twice the natural roll frequency can bring the vessel dynamics into a bifurcation mode and lead to extreme values of roll. Recent years have seen several incidents with dramatic damage to container vessels...... the ship's speed and course, to escape from the bifurcation condition. This paper proposes nonparametric methods to detect the onset of roll resonance and demonstrates their performance. Theoretical conditions for parametric resonance are revisited and are used to develop efficient methods to detect its...... on experimental data from model tests and on data from a container ship crossing the Atlantic during a storm....

Prestressed concrete reactor vessel thermal cylinder model study

International Nuclear Information System (INIS)

Callahan, J.P.; Canonico, D.A.; Richardson, M.; Corum, J.M.; Dodge, W.G.; Robinson, G.C.; Whitman, G.D.

1977-01-01

The thermal cylinder experiment was designed both to provide information for evaluating the capability of analytical methods to predict the time-dependent stress-strain behavior of a 1 / 6 -scale model of the barrel section of a single-cavity prestressed concrete reactor vessel and to demonstrate the structural behavior under design and off-design thermal conditions. The model was a thick-walled cylinder having a height of 1.22 m, a thickness of 0.46 m, and an outer diameter of 2.06 m. It was prestressed both axially and circumferentially and subjected to 4.83 MPa internal pressure together with a thermal crossfall imposed by heating the inner surface to 338.8 K and cooling the outer surface to 297.1 K. The initial 460 days of testing were divided into time periods that simulated prestressing, heatup, reactor operation, and shutdown. At the conclusion of the simulated operating period, the model was repressurized and subjected to localized heating at 505.4 K for 84 days to produce an off-design hot-spot condition. Comparisons of experimental data with calculated values obtained using the SAFE-CRACK finite-element computer program showed that the program was capable of predicting time-dependent behavior in a vessel subjected to normal operating conditions, but that it was unable to accurately predict the behavior during off-design hot-spot heating. Readings made using a neutron and gamma-ray backscattering moisture probe showed little, if any, migration of moisture in the concrete cross section. Destructive examination indicated that the model maintained its basic structural integrity during localized hot-spot heating

Combining operational models and data into a dynamic vessel risk assessment tool for coastal regions

Science.gov (United States)

Fernandes, R.; Braunschweig, F.; Lourenço, F.; Neves, R.

2016-02-01

The technological evolution in terms of computational capacity, data acquisition systems, numerical modelling and operational oceanography is supplying opportunities for designing and building holistic approaches and complex tools for newer and more efficient management (planning, prevention and response) of coastal water pollution risk events. A combined methodology to dynamically estimate time and space variable individual vessel accident risk levels and shoreline contamination risk from ships has been developed, integrating numerical metocean forecasts and oil spill simulations with vessel tracking automatic identification systems (AIS). The risk rating combines the likelihood of an oil spill occurring from a vessel navigating in a study area - the Portuguese continental shelf - with the assessed consequences to the shoreline. The spill likelihood is based on dynamic marine weather conditions and statistical information from previous accidents. The shoreline consequences reflect the virtual spilled oil amount reaching shoreline and its environmental and socio-economic vulnerabilities. The oil reaching shoreline is quantified with an oil spill fate and behaviour model running multiple virtual spills from vessels along time, or as an alternative, a correction factor based on vessel distance from coast. Shoreline risks can be computed in real time or from previously obtained data. Results show the ability of the proposed methodology to estimate the risk properly sensitive to dynamic metocean conditions and to oil transport behaviour. The integration of meteo-oceanic + oil spill models with coastal vulnerability and AIS data in the quantification of risk enhances the maritime situational awareness and the decision support model, providing a more realistic approach in the assessment of shoreline impacts. The risk assessment from historical data can help finding typical risk patterns ("hot spots") or developing sensitivity analysis to specific conditions, whereas real

System for cooling the upper wall of a nuclear reactor vessel

International Nuclear Information System (INIS)

Pailla, Henri; Schaller, Karl; Vidard, Michel.

1974-01-01

A system for cooling the upper wall of the main vessel of a fast neutron reactor is described. This vessel is suspended from an upper shield by the upper wall. It includes coils carrying a coolant which are immersed in an intermediate liquid bathing the wall and contained in a tank integral with the vessel. At least one of the two cooling and intermediate liquids is a liquid metal. The main vessel is contained in a safety vessel, the space between the main and safety vessels is occluded in its upper part by an insulating shield placed under the tank. There is a liquid metal seal between the upper wall and the upper shield under the tank. This system has been specially designed for sodium cooled fast neutron reactors [fr

Coastal Logbook Survey (Vessels)

Data.gov (United States)

National Oceanic and Atmospheric Administration, Department of Commerce — This data set contains catch (landed catch) and effort for fishing trips made by vessels that have been issued a Federal permit for the Gulf of Mexico reef fish,...

Comparison of a networks-of-zones fluid mixing model for a baffled stirred vessel with three-dimensional electrical resistance tomography

International Nuclear Information System (INIS)

Rodgers, T L; Siperstein, F R; Mann, R; York, T A; Kowalski, A

2011-01-01

Reliable models for the simulation of mixing vessels are important for the understanding of real-life mixing problems. To achieve these models, information about the mixing in the system must be measured to compare with the predicted values. Electrical resistance tomography has the capability to measure spatial and temporal changes within a vessel in three dimensions even in optically inaccessible environments. This paper discusses the creation of a network-of-zones model for the prediction of mixing within a vessel with a Cowles disc-type agitator. Solving of the network-of-zones simplified transport equations for the vessel predicts the concentration distribution of an inert tracer added to the vessel. The change in this distribution with time is calculated and compared with visual inspection of the vessel. The concentration distribution inside the vessel is also measured using electrical resistance tomography and shows good agreement with the predicted values

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

Probabilistic retinal vessel segmentation

Science.gov (United States)

Wu, Chang-Hua; Agam, Gady

2007-03-01

Optic fundus assessment is widely used for diagnosing vascular and non-vascular pathology. Inspection of the retinal vasculature may reveal hypertension, diabetes, arteriosclerosis, cardiovascular disease and stroke. Due to various imaging conditions retinal images may be degraded. Consequently, the enhancement of such images and vessels in them is an important task with direct clinical applications. We propose a novel technique for vessel enhancement in retinal images that is capable of enhancing vessel junctions in addition to linear vessel segments. This is an extension of vessel filters we have previously developed for vessel enhancement in thoracic CT scans. The proposed approach is based on probabilistic models which can discern vessels and junctions. Evaluation shows the proposed filter is better than several known techniques and is comparable to the state of the art when evaluated on a standard dataset. A ridge-based vessel tracking process is applied on the enhanced image to demonstrate the effectiveness of the enhancement filter.

Design and analysis of prestressed reactor vessels

International Nuclear Information System (INIS)

Burrow, R.E.D.

1978-01-01

This review is intended to draw attention to subjects of interest from papers given at two sessions of the SMiRT 4 conference. The first of these is the structural engineering of prestressed reactor vessels. The topics include developments in the general design of prestressed vessels, structural analysis of PCVRs, model tests and design of penetration, closures and liners for PCVRs. The question of gas cracks was amongst other issues raised. The second of the sessions was concerned with loading conditions and structural analysis of reactor containment. Reference is made to a variety of topics discussed in this session. Particular attention is given to the effects caused by missiles. In concluding, the reviewer suggests the need for a critical assessment of the existing mass of information to sort out the essentials and to bring back some simplicity into design analysis. (UK)

Rates of chemical reaction and atmospheric heating during core debris expulsion from a pressurized vessel

International Nuclear Information System (INIS)

Powers, D.A.; Tarbell, W.W.; Brockman, J.E.; Pilch, M.

1986-01-01

Core debris may be expelled from a pressurized reactor vessel during a severe nuclear reactor accident. Experimental studies of core debris expulsion from pressurized vessels have established that the expelled material can be lofted into the atmosphere of the reactor containment as particulate 0.4 to 2 mm in diameter. These particles will vigorously react with steam and oxygen in the containment atmosphere. Data on such reactions during tests with 80 kg of expelled melt will be reported. A model of the reaction rates based on gas phase mass transport will be described and shown to account for atmospheric heating and aerosol generation observed in the tests

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.)

Computational Fluid Dynamics Analysis of Pulsatile Blood Flow Behavior in Modelled Stenosed Vessels with Different Severities

Directory of Open Access Journals (Sweden)

Mohsen Mehrabi

2012-01-01

Full Text Available This study focuses on the behavior of blood flow in the stenosed vessels. Blood is modelled as an incompressible non-Newtonian fluid which is based on the power law viscosity model. A numerical technique based on the finite difference method is developed to simulate the blood flow taking into account the transient periodic behaviour of the blood flow in cardiac cycles. Also, pulsatile blood flow in the stenosed vessel is based on the Womersley model, and fluid flow in the lumen region is governed by the continuity equation and the Navier-Stokes equations. In this study, the stenosis shape is cosine by using Tu and Devil model. Comparing the results obtained from three stenosed vessels with 30%, 50%, and 75% area severity, we find that higher percent-area severity of stenosis leads to higher extrapressure jumps and higher blood speeds around the stenosis site. Also, we observe that the size of the stenosis in stenosed vessels does influence the blood flow. A little change on the cross-sectional value makes vast change on the blood flow rate. This simulation helps the people working in the field of physiological fluid dynamics as well as the medical practitioners.

Fission product aerosol removal test by containment spray under accident management conditions (3)

International Nuclear Information System (INIS)

Watanabe, Atsushi; Nagasaka, Hideo; Yokobori, Seiichi; Akinaga, Makoto

2000-01-01

In order to demonstrate the effective FP aerosol removal by containment spray under Japanese AM conditions, two system integral tests and two separate effect tests were carried out using a full-height simulation test facility. In case of PWR LOCA, aerosol concentration in the upper containment vessel decreased even under low spray flow rate. In case of BWR LOCA with water injection into RPV, the aerosol concentration in the entire vessel also decreased rapidly after aerosol supply stopping. In both cases, the removal rate estimated from the NUREG-1465 was coincided with test results. The aerosol washing effect by spray was confirmed to be predominant by conducting suppression chamber isolation test. It turned out that the effect of aerosol solubility and density on aerosol removal by spray was quite small by conducting insoluble aerosol injection test. After the modification of aerosol removal model by the spray and hygroscopic aerosol model in original MELCOR 1.8.4, calculated aerosol concentration transient in the containment vessel agreed well with the test data. (author)

Adaptable three-dimensional Monte Carlo modeling of imaged blood vessels in skin

Science.gov (United States)

Pfefer, T. Joshua; Barton, Jennifer K.; Chan, Eric K.; Ducros, Mathieu G.; Sorg, Brian S.; Milner, Thomas E.; Nelson, J. Stuart; Welch, Ashley J.

1997-06-01

In order to reach a higher level of accuracy in simulation of port wine stain treatment, we propose to discard the typical layered geometry and cylindrical blood vessel assumptions made in optical models and use imaging techniques to define actual tissue geometry. Two main additions to the typical 3D, weighted photon, variable step size Monte Carlo routine were necessary to achieve this goal. First, optical low coherence reflectometry (OLCR) images of rat skin were used to specify a 3D material array, with each entry assigned a label to represent the type of tissue in that particular voxel. Second, the Monte Carlo algorithm was altered so that when a photon crosses into a new voxel, the remaining path length is recalculated using the new optical properties, as specified by the material array. The model has shown good agreement with data from the literature. Monte Carlo simulations using OLCR images of asymmetrically curved blood vessels show various effects such as shading, scattering-induced peaks at vessel surfaces, and directionality-induced gradients in energy deposition. In conclusion, this augmentation of the Monte Carlo method can accurately simulate light transport for a wide variety of nonhomogeneous tissue geometries.

Structural features and in-service inspection of the LTHR-200 pressure vessel

International Nuclear Information System (INIS)

Xiong Dunshi; He Shuyan; Liu Junjie; Yu Suyuan

1993-01-01

LTHR-200 is a low temperature district-heating reactor. It adopts double-shell design pressure vessel and metal containment. Because of the safety and structural features of the reactor, the in-service inspection of the pressure vessel can be simplified greatly. LTHR-200 is an integrated arrangement. Both its core components and the main heat exchangers are contained in the reactor pressure vessel. The coolant of the main loop is run by a full-power natural circulation and there need no main pumps and pipes. Thus, the reactor pressure vessel constitutes the pressure boundary of the reactor's main loop coolant. In regard to these features, a small-sized containment is designed for the reactor. The metal safety container with a small volume is placed closely around the reactor pressure vessel. Outside the metal containment, there is a large reinforced concrete construction for the reactor. Their main operation and design parameters are as follows: The pressure vessel: operation pressure = 2.4 MPa; design pressure = 3.0 MPa; design temperature = 250 deg C; 40 year fast neutron (E>1MeV) fluence in the belt-line region = < 10E16n/cm; internal diameter = 5000 mm; material SA516-70; shell thickness 65 mm; The metal containment: maximum operation pressure = 1.8 MPa; design pressure = 1.8 MPa; design temperature = 250 deg. C; upper internal diameter 7000 mm; lower internal diameter = 5600 mm; material = SA516-70; shell thickness, upper part = 80 mm; lower part = 50 mm. All penetrating pipes through the pressure vessel are located at the top penetration section of the shell. All the internal diameters of penetrating pipes are less than 50 mm. Inside and outside the metal containment wall respectively, isolating valves are connected to the reactor coolant pipe which passes through the containment. These two isolating valves use different driving methods. Every penetrating part of the reactor construction uses a proper form of structure according to safety requirements

Method of producing the arched surfaces of diaphragm rings for large containers, especially for prestressed-concrete pressure vessels of nuclear reactors

International Nuclear Information System (INIS)

Kumpf, H.

1976-01-01

In producing arched surfaces of diaphragm rings for large containers, especially for prestressed-concrete pressure vessels for nuclear power plants, it is of advantage to manufacture these directly on the construction site. According to the invention the, at first level, diaphragm ring is put on the predetermined place, sectionally pressed against and shaped by a shaping tool - with a profiled supporting ring as a counter-acting tool - and afterwards welded together with the annular wall sections of the large container along the shaped parts. The manufacture of single and double configurations of diaphragm rings is described. It is of advantage if shaping and mounting position coincide. (UWI) [de

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

Development of mobile manipulator for maintenance work in containment vessels of nuclear power plants

International Nuclear Information System (INIS)

Omichi, Takeo; Nishihara, Masatoshi; Hosaka, Shigetaka; Nakayama, Junji; Sato, Masatoshi; Ishida, Michiyasu

1985-01-01

The teleoperation system with robot is described for in the containment vessels of nuclear power plants. We have developed a high level robot system as the practical use level. The robot is designed to execute the locomotions and manipulations required for closing and opening the valve, tightening the bolt and others. The robot consists of a locomotor with four legs and two driving wheels, an articulated manipulator with seven joints, and an ITV arm with stereo-camera. The size of the robot is small, that is about 500 mm in length, 500 mm in width, 1200 mm in height and 420 kg in weight. The robot can be operated in a hostile environment, which has a 10 6 R gamma ray dose, 70 deg C temperature, 100 % relative humidity. We have added an advanced control method in order to reduce the operator's load. Also, an interlock and a fail-safe control are installed in the robot system. (author)

Loads on EPR containment after RPV failure at high pressure; Belastungen des EPR-Containments in Falle eines RDB-Versagens bei hohem Druck

Energy Technology Data Exchange (ETDEWEB)

Jacobs, G.

1995-08-01

As regards the desgin of the EPR, the general strategy is to eliminate, the vessel failure at high pressure by preventive and mitigative measures. The design proposals involved trust in the reliability of dedicated devices (relief valves) for rapid depressurization. The aim is to attain a lower pressure level at the moment of vessel failure, so that the containment is capable to cope with the blowdown impact on the pit walls and the vessel supporting structures. Nevertheless, the potential of a high-pressure failure of the vessel must be kept in mind, whatever well thought-out and reliable preventive depressurization measures might be. Therefore, the reactor pressure blowdown has been studied in order to quantify the ultimate containment load, which might support future design requirements. The calculations were performed with the LWR transient analysis thermal-hydraulics computer code REALAP5/MOD3. In previous analyses, the nodalization of the problem was based on the geometrical conditions of a typical German 1300 MW(e) NPP. In the present analysis a new input model has been used, which was based on the EPR conditions. (orig./HP)

Polynomial fuzzy model-based approach for underactuated surface vessels

DEFF Research Database (Denmark)

Khooban, Mohammad Hassan; Vafamand, Navid; Dragicevic, Tomislav

2018-01-01

The main goal of this study is to introduce a new polynomial fuzzy model-based structure for a class of marine systems with non-linear and polynomial dynamics. The suggested technique relies on a polynomial Takagi–Sugeno (T–S) fuzzy modelling, a polynomial dynamic parallel distributed compensation...... surface vessel (USV). Additionally, in order to overcome the USV control challenges, including the USV un-modelled dynamics, complex nonlinear dynamics, external disturbances and parameter uncertainties, the polynomial fuzzy model representation is adopted. Moreover, the USV-based control structure...... and a sum-of-squares (SOS) decomposition. The new proposed approach is a generalisation of the standard T–S fuzzy models and linear matrix inequality which indicated its effectiveness in decreasing the tracking time and increasing the efficiency of the robust tracking control problem for an underactuated...

Experimental and theoretical studies on the high pressure vessel

International Nuclear Information System (INIS)

So, Dong Sup

1992-02-01

A High Pressure Melt Ejection (HPME) is one of the most important phenomena relevant to Direct Containment Heating(DCH) which could lead to an early containment failure in a several accident of PWRs. Dispersal of core debris following a postulated high pressure failure of PWR reactor vessel has been investigated by experimental works and one-dimensional computer modeling to find the relation between the fraction of melt simulant retained in the cavity and the reactor vessel initial conditions as well as to examine the hydrodynamic processes in a reactor cavity geometry. Simulated HPME experiments have been performed with two small-scale (1/25-th and 1/41-st) transparent reactor cavity models of the Young-Gwang unit 1 and 2. Wood's metal and water have been used as melt sumulants while high pressure nitrogen and carbon dioxide have been used as driver gases to simulate the blowdown steam and gas from the breach of the reactor pressure vessel. The high speed movies of the transient tests showed that no fraction of the melt simulant exits the cavity model via the vertical cavity tunnel under its own momentum, and that the discharged simulant from the pressure vessel exits the reactor cavity model during the gas blowdown. The principal removal mechanism seemed to be a combined mechanism of film entrainment and particle levitation due to the driving force of the blowdown gas. Experimental data for the fraction of melt simulant retained in the cavity model (Y f ) during a postulated scenario of the HPME from PWR pressure vessels have been obtained as a function of various test parameters. These data have been used to develop a correlation for Y f that fits all the data (a total of 313 data points) within the standard deviation of 0.054 by means of dimensional analysis and nonlinear least squares optimization technique. The basic effects of important parameters used to describe the HPME accident sequence on the Y f are determined based on the correlation obtained here and

TPX vacuum vessel transient thermal and stress conditions

International Nuclear Information System (INIS)

Feldshteyn, Y.; Dinkevich, S.; Feng, T.; Majumder, D.

1995-01-01

The TPX vacuum vessel provides the vacuum boundary for the plasma and the mechanical support for the internal components. Another function of the vacuum vessel is to contain neutron shielding water in the double wall space during normal operation. This double wall space serves as a heat reservoir for the entire vacuum vessel during bakeout. The vacuum vessel and the internal components are subjected to thermal stresses induced by a nonuniform temperature distribution within the structure during bakeout. A successful Conceptual Design Review in March 1993 has established superheated steam as the heating source of the vacuum vessel. A transient bakeout mode of the vacuum vessel and in-vessel components has been analyzed to evaluate transient period duration, proper temperature level, actual thermal stresses and performance of the steam equipment. Thermally, the vacuum vessel structure may be considered as an adiabatic system because it is perfectly insulated by the strong surrounding vacuum and multiple layers of superinsulation. Important aspects of the analysis are described herein

Structural Analysis of the NCSX Vacuum Vessel

International Nuclear Information System (INIS)

Fred Dahlgren; Art Brooks; Paul Goranson; Mike Cole; Peter Titus

2004-01-01

The NCSX (National Compact Stellarator Experiment) vacuum vessel has a rather unique shape being very closely coupled topologically to the three-fold stellarator symmetry of the plasma it contains. This shape does not permit the use of the common forms of pressure vessel analysis and necessitates the reliance on finite element analysis. The current paper describes the NCSX vacuum vessel stress analysis including external pressure, thermal, and electro-magnetic loading from internal plasma disruptions and bakeout temperatures of up to 400 degrees centigrade. Buckling and dynamic loading conditions are also considered

Fission reactor container

International Nuclear Information System (INIS)

Akagawa, Katsuhiko.

1991-01-01

Cooling water is sent without using dynamic equipments upon loss of coolants accident in a pressure vessel by improving an arrangement of a nuclear reactor pressure vessel. That is, a containing space is formed at the center of a suppression chamber for storing cooling water while being partitioned with each other, in which the pressure vessel is placed. Further, a water reservoir is formed above the pressure vessel. Then a water discharge pipe is connected to the reservoir for submerging the stored water over the pressure vessel upon occurrence of loss of coolants accident. Further, a water injection pipe is disposed between the pressure suppression chamber and the pressure vessel for injecting the cooling water in the pressure suppression chamber to the reactor core of the pressure vessel by the difference of a water head upon loss of coolants accident. With such a constitution, the pressure vessel has high earthquake proofness. Further, upon loss of coolants accident of the pressure vessel, the cooling water in the reservoir is discharged to submerge and cool the pressure vessel efficiently. Further, the reactor core of the pressure vessel can certainly be cooled by the cooling water of the pressure suppression chamber without relying on dynamic equipments. (I.S.)

Pressure vessel lid

International Nuclear Information System (INIS)

Schoening, J.; Elter, C.; Becker, G.; Pertiller, S.

1986-01-01

The invention concerns a lid for closing openings in reactor pressure vessels containing helium, which is made as a circular casting with hollow spaces and a flat floor and is set on the opening and kept down. It consists of helium-tight metal cast material with sufficient temperature resistance. There are at least two concentric heat resistant seals let into the bottom of the lid. The bottom is in immediate contact with the container atmosphere and has hollow spaces in its inside in the area opposite to the opening. (orig./HP) [de

Neutron irradiation effects in reactor pressure vessel steels and weldments. Working document

International Nuclear Information System (INIS)

1998-10-01

As a result of the popularity of the Agencies report 'Neutron Irradiation Embrittlement of Reactor Pressure Vessel Steels' of 1975, it was decided that another report on this broad subject would be of use. In this report, background and contemporary views on specially identified areas of the subject are considered as self-contained chapters, written by experts. A separate abstract was prepared for the introduction and for each of the eleven chapters, which are: 1. Reactor Pressure Vessel Design, 2. Reactor Pressure Materials, 3. WWER Pressure Vessels, 4. Determination of Mechanical Properties, 5. Neutron Exposure, 6. Methodology of Irradiation Experiments, 7. Effect of Irradiation on Mechanical Properties, 8. Mechanisms of Irradiation Embrittlement, 9. Modelling of Irradiation Damage, 10. Annealing of Irradiation Damage, 11. Safety Assessment using Surveillance Programmes and Data Bases

Analytical investigation of multicavity prestressed concrete pressure vessels for elastic loading conditions

International Nuclear Information System (INIS)

Fanning, D.N.

1978-09-01

A three-dimensional finite-element analysis of a commercial high-temperature gas-cooled reactor (HTGR) was made using the finite-element code STATIC-SAP. Four loading conditions were analyzed elastically to evaluate the behavior of the concentric core prestressed concrete reactor vessel (PCRV) of the HTGR. The results of the analysis were evaluated in accordance with Section III, Division 2, of the ASME Code for Reactor Vessels and Containments. The calculated maximum stresses were found to be well within the Code-allowable values. The analysis was preceded by an evaluation of candidate computer codes using comparisons of experimental data with analytical results for the Ohbayashi-Gumi multicavity PCRV model. This vessel was chosen as a basis for comparison because of its geometrical similarity to the large multicavity PCRV and the anticipated availability of a complete set of the original experimental data. The three-dimensional finite-element codes NONSAP and STATIC-SAP were used for the analysis of the Ohbayashi-Gumi vessel

46 CFR 176.812 - Pressure vessels and boilers.

Science.gov (United States)

2010-10-01

... 46 Shipping 7 2010-10-01 2010-10-01 false Pressure vessels and boilers. 176.812 Section 176.812... TONS) INSPECTION AND CERTIFICATION Material Inspections § 176.812 Pressure vessels and boilers. (a.... (b) Periodic inspection and testing requirements for boilers are contained in § 61.05 in subchapter F...

Prestressed concrete pressure vessels for nuclear reactors - 1973

Energy Technology Data Exchange (ETDEWEB)

1977-01-01

This standard deals with the design, construction, inspection and testing of prestressed concrete pressure vessels for nuclear reactors. Such pressure vessels serve the dual purpose of shielding and containing gas cooled nuclear reactors and are a form of civil engineering structure requiring particularly high integrity, and ensured leak tightness. (Metric)

Prenatal Experiences of Containment in the Light of Bion's Model of Container/Contained

Science.gov (United States)

Maiello, Suzanne

2012-01-01

This paper explores the idea of possible proto-experiences of the prenatal child in the context of Bion's model of container/contained. The physical configuration of the embryo/foetus contained in the maternal uterus represents the starting point for an enquiry into the unborn child's possible experiences of its state of being contained in a…

Predictability of steel containment response near failure track 3 - structural integrity, dynamic behavior, and seismic design

International Nuclear Information System (INIS)

Costello, J.F.; Ludwigsen, J.S.; Luk, V.K.; Hessheimer, M.F.

2000-01-01

The Nuclear Power Engineering Corporation of Japan and the US Nuclear Regulatory Commission Office of Nuclear Regulatory Research, are co-sponsoring and jointly funding a Cooperative Containment Research Program at Sandia National Laboratories, Albuquerque, New Mexico, USA. As a part of this program, a steel containment vessel model and contact structure assembly was tested to failure with over pressurization at Sandia on December 11--12, 1996. The steel containment vessel model was a mixed-scale model (1:10 in geometry and 1:4 in shell thickness) of a steel containment for an improved Mark-II Boiling Water Reactor plant in Japan. The contact structure, which is a thick, bell-shaped steel shell separated at a nominally uniform distance from the model, provides a simplified representation of features of the concrete reactor shield building in the actual plant. The objective of the internal pressurization test was to provide measurement data of the structural response of the model up to its failure in order to validate analytical modeling, to find its pressure capacity, and to observe the failure model and mechanisms

Design and analysis of reactor containment of steel-concrete composite laminated shell

International Nuclear Information System (INIS)

Ichikawa, K.

1977-01-01

Reinforced and prestressed concrete containments for reactors have been developed in order to avoid the difficulties of welding of steel containments encountered as their capacities have become large: growing thickness of steel shells gave rise to the requirement of stress relief at the construction sites. However, these concrete vessels also seem to face another difficulty: the lack of shearing resistance capacity. In order to improve the shearing resistance capacity of the containment vessel, while avoiding the difficulty of welding, a new scheme of containment consisting of steel-concrete laminated shell is being developed. In the main part of a cylindrical vessel, the shell consists of two layers of thin steel plates located at the inner and outer surfaces, and a layer of concrete core into which both the steel plates are anchored. In order to validate the feasibility and safety of this new design, the results of analysis on the basis of up-to-date design loads are presented. The results of model tests in 1:30 scale are also reported. (Auth.)

Computational study of the mixed cooling effects on the in-vessel retention of a molten pool in a nuclear reactor

International Nuclear Information System (INIS)

Kim, Byung Seok; Sohn, Chang Hyun; Ahn, Kwang Il

2004-01-01

The retention of a molten pool vessel cooled by internal vessel reflooding and/or external vessel reactor cavity flooding has been considered as one of severe accident management strategies. The present numerical study investigates the effect of both internal and external vessel mixed cooling on an internally heated molten pool. The molten pool is confined in a hemispherical vessel with reference to the thermal behavior of the vessel wall. In this study, our numerical model used a scaled-down reactor vessel of a KSNP (Korea Standard Nuclear Power) reactor design of 1000 MWe (a pressurized water reactor with a large and dry containment). Well-known temperature-dependent boiling heat transfer curves are applied to the internal and external vessel cooling boundaries. Radiative heat transfer has been considered in the case of dry internal vessel boundary condition. Computational results show that the external cooling vessel boundary conditions have better effectiveness than internal vessel cooling in the retention of the melt pool vessel failure

A mathematical model for batch and continuous thickening of flocculent suspensions in vessels with varying section

Energy Technology Data Exchange (ETDEWEB)

Buerger, R.; Damasceno, J.J.R.; Karlesen, K.H.

2001-10-01

The phenomenological theory of continuous thickening of flocculated suspensions in an ideal cylindrical thickener is extended to vessels having varying cross-section, including divergent or convergent conical vessels. The purpose of this contribution is to draw attention to the corresponding mathematical model, whose key ingredient is a strongly degenerate parabolic partial differential equation. For ideal (non-flocculated) suspensions, which do not form co compressible sediments, the mathematical model reduces to the kinematic approach by Anestis, who developed a method of construction of exact solution by the method of characteristics. The difficulty lies in the fact that characteristics and iso-concentration lines, unlike the conventional Kynch model for cylindrical vessels, do not coincide, and one has to resort to numerical methods to simulate the thickening process. A numerical algorithm is presented and employed for simulations of continuous thickening. Implications of the mathematical model are also demonstrated by steady-state calculations, which lead to new possibilities in thickener design. (author)

19 CFR 4.7d - Container status messages.

Science.gov (United States)

2010-04-01

... 19 Customs Duties 1 2010-04-01 2010-04-01 false Container status messages. 4.7d Section 4.7d... TREASURY VESSELS IN FOREIGN AND DOMESTIC TRADES Arrival and Entry of Vessels § 4.7d Container status messages. (a) Container status messages required. In addition to the advance filing requirements pursuant...

Fuel containing vessel for transporting nuclear fuel

International Nuclear Information System (INIS)

Yoshizawa, Hiroyasu; Shimizu, Fukuzo; Tanaka, Nobuyuki.

1996-01-01

A shock absorbing mechanism is disposed on an inner bottom of a vessel main body. The shock absorbing mechanism comprises a shock absorbing member disposed on the upper surface of a bottom wall, an annular metal plate disposed on the upper surface of the shock absorbing member and an annular spacer disposed on the upper surface of the metal plate. The shock absorbing member is made of a material such as of wood, lead, metal honeycomb or a metal mesh, which plastically deforms when applied with load higher than a predetermined level, and is formed in a square block-like form covering the upper surface of the bottom wall. The spacer is made of a thin soft material such as tetrafluoroethylene, and is formed in such a shape as capable of preventing direct contact of the lower end of the cylindrical member in a lower tie plate of nuclear fuels with the metal portion. This can ensure integrity of nuclear fuels even when they fall from a high place upon an assumed dropping accident. (I.N.)

Modeling of melt retention in EU-APR1400 ex-vessel core catcher

Energy Technology Data Exchange (ETDEWEB)

Granovsky, V. S.; Sulatsky, A. A.; Khabensky, V. B.; Sulatskaya, M. B. [Alexandrov Research Inst. of Technology NITI, Sosnovy Bor (Russian Federation); Gusarov, V. V.; Almyashev, V. I.; Komlev, A. A. [Saint Petersburg State Technological Univ. SPbSTU, St.Petersburg (Russian Federation); Bechta, S. [KTH, Stockholm (Sweden); Kim, Y. S. [KHNP, 1312 Gil 70, Yuseongdaero, Yuseong-gu, Daejeon (Korea, Republic of); Park, R. J.; Kim, H. Y.; Song, J. H. [KAERI, 989 Gil 111, Daedeokdaero, Yuseong-gu, Daejeon (Korea, Republic of)

2012-07-01

A core catcher is adopted in the EU-APR1400 reactor design for management and mitigation of severe accidents with reactor core melting. The core catcher concept incorporates a number of engineering solutions used in the catcher designs of European EPR and Russian WER-1000 reactors, such as thin-layer corium spreading for better cooling, retention of the melt in a water-cooled steel vessel, and use of sacrificial material (SM) to control the melt properties. SM is one of the key elements of the catcher design and its performance is critical for melt retention efficiency. This SM consists of oxide components, but the core catcher also includes sacrificial steel which reacts with the metal melt of the molten corium to reduce its temperature. The paper describes the required properties of SM. The melt retention capability of the core catcher can be confirmed by modeling the heat fluxes to the catcher vessel to show that it will not fail. The fulfillment of this requirement is demonstrated on the example of LBLOCA severe accident. Thermal and physicochemical interactions between the oxide and metal melts, interactions of the melts with SM, sacrificial steel and vessel, core catcher external cooling by water and release of non-condensable gases are modeled. (authors)

Containment for low temperature district nuclear-heating reactor

International Nuclear Information System (INIS)

He Shuyan; Dong Duo

1992-03-01

Integral arrangement is adopted for Low Temperature District Nuclear-heating Reactor. Primary heat exchangers, control rod drives and spent fuel elements are put in the reactor pressure vessel together with reactor core. Primary coolant flows through reactor core and primary heat exchangers in natural circulation. Primary coolant pipes penetrating the wall of reactor pressure vessel are all of small diameters. The reactor vessel constitutes the main part of pressure boundary of primary coolant. Therefore the small sized metallic containment closed to the wall of reactor vessel can be used for the reactor. Design principles and functions of the containment are as same as the containment for PWR. But the adoption of small sized containment brings about some benefits such as short period of manufacturing, relatively low cost, and easy for sealing. Loss of primary coolant accident would not be happened during the rupture accident of primary coolant pressure boundary inside the containment owing to its intrinsic safety

Modeling and analysis of alternative concept of ITER vacuum vessel primary heat transfer system

International Nuclear Information System (INIS)

Carbajo, Juan; Yoder, Graydon; Dell'Orco, G.; Curd, Warren; Kim, Seokho

2010-01-01

A RELAP5-3D model of the ITER (Latin for 'the way') vacuum vessel (VV) primary heat transfer system has been developed to evaluate a proposed design change that relocates the heat exchangers (HXs) from the exterior of the tokamak building to the interior. This alternative design protects the HXs from external hazards such as wind, tornado, and aircraft crash. The proposed design integrates the VV HXs into a VV pressure suppression system (VVPSS) tank that contains water to condense vapour in case of a leak into the plasma chamber. The proposal is to also use this water as the ultimate sink when removing decay heat from the VV system. The RELAP5-3D model has been run under normal operating and abnormal (decay heat) conditions. Results indicate that this alternative design is feasible, with no effects on the VVPSS tank under normal operation and with tank temperature and pressure increasing under decay heat conditions resulting in a requirement to remove steam generated if the VVPSS tank low pressure must be maintained.

Investigation of vessel exterior air cooling for a HLMC reactor

International Nuclear Information System (INIS)

Sienicki, J. J.; Spencer, B. W.

2000-01-01

The Secure Transportable Autonomous Reactor (STAR) concept under development at Argonne National Laboratory provides a small (300 MWt) reactor module for steam supply that incorporates design features to attain proliferation resistance, heightened passive safety, and improved cost competitiveness through extreme simplification. Examples are the achievement of 100%+ natural circulation heat removal from the low power density/low pressure drop ultra-long lifetime core and utilization of lead-bismuth eutectic (LBE) coolant enabling elimination of main coolant pumps as well as the need for an intermediate heat transport circuit. It is required to provide a passive means of removing decay heat and effecting reactor cooldown in the event that the normal steam generator heat sink, including its normal shutdown heat removal mode, is postulated to be unavailable. In the present approach, denoted as the Reactor Exterior Cooling System (RECS), passive decay heat removal is provided by cooling the outside of the containment/guard vessel with air. RECS is similar to the Reactor Vessel Auxiliary Cooling System (RVACS) incorporated into the PRISM design. However, to enhance the heat removal, RECS incorporates fins on the containment vessel exterior to enhance heat transfer to air as well as removable steel venetian conductors that provide a conduction heat transfer path across the reactor vessel-containment vessel gap to enhance heat transfer between the vessels. The objective of the present work is to investigate the effectiveness of air cooling in removing heat from the vessel and limiting the coolant temperature increase following a sudden complete loss of the steam generator heat sink

Modelling of hydrogen deflagration in a vented vessel

International Nuclear Information System (INIS)

Wang, L.L.; Wong, R.C.

1995-01-01

Hydrogen Deflagration inside closed and vented 2.3 m diameter vessels were simulated by using the GOTHIC lumped-parameter computer code. Different cell arrangements were used in the modelling. Other parameters such as flame speed and hydrogen concentration were studied. It was found that the calculated peak pressures for cases using the experimental measured burn durations were close to the pressures measured from the experiments. When the default flame speed was used, higher peak pressure was predicted by GOTHIC. This could be explained by the the fact that the default flame speed used in the GOTHIC burn model was based on the results of a large scale test with moderate turbulence level. However, the overall results of the pressure transients were comparable with the experimental data. In addition, time and spatial convergencies of the model were also studied. The peak pressure estimated by modelling the sphere as five or more spherical cells was shown to converge to within +/- 3 percent. (author). 8 refs., 6 tabs., 9 figs

Coastal Discard Logbook Survey (Vessels)

Data.gov (United States)

National Oceanic and Atmospheric Administration, Department of Commerce — This data set contains data on the type and amount of marine resources that are discarded or interacted with by vessels that are selected to report to the Southeast...

Development of a Weibull model of cleavage fracture toughness for shallow flaws in reactor pressure vessel material

Energy Technology Data Exchange (ETDEWEB)

Bass, B.R.; Williams, P.T.; McAfee, W.J.; Pugh, C.E. [Oak Ridge National Lab., Heavy-Section Steel Technology Program, Oak Ridge, TN (United States)

2001-07-01

A primary objective of the United States Nuclear Regulatory Commission (USNRC) -sponsored Heavy-Section Steel Technology (HSST) Program is to develop and validate technology applicable to quantitative assessments of fracture prevention margins in nuclear reactor pressure vessels (RPVs) containing flaws and subjected to service-induced material toughness degradation. This paper describes an experimental/analytical program for the development of a Weibull statistical model of cleavage fracture toughness for applications to shallow surface-breaking and embedded flaws in RPV materials subjected to multi-axial loading conditions. The experimental part includes both material characterization testing and larger fracture toughness experiments conducted using a special-purpose cruciform beam specimen developed by Oak Ridge National Laboratory for applying biaxial loads to shallow cracks. Test materials (pressure vessel steels) included plate product forms (conforming to ASTM A533 Grade B Class 1 specifications) and shell segments procured from a pressurized-water reactor vessel intended for a nuclear power plant. Results from tests performed on cruciform specimens demonstrated that biaxial loading can have a pronounced effect on shallow-flaw fracture toughness in the lower-transition temperature region. A local approach methodology based on a three-parameter Weibull model was developed to correlate these experimentally-observed biaxial effects on fracture toughness. The Weibull model, combined with a new hydrostatic stress criterion in place of the more commonly used maximum principal stress in the kernel of the Weibull stress integral definition, is shown to provide a scaling mechanism between uniaxial and biaxial loading states for 2-dimensional flaws located in the A533-B plate material. The Weibull stress density was introduced as a matrice for identifying regions along a semi-elliptical flaw front that have a higher probability of cleavage initiation. Cumulative

Development of a Weibull model of cleavage fracture toughness for shallow flaws in reactor pressure vessel material

International Nuclear Information System (INIS)

Bass, B.R.; Williams, P.T.; McAfee, W.J.; Pugh, C.E.

2001-01-01

A primary objective of the United States Nuclear Regulatory Commission (USNRC) -sponsored Heavy-Section Steel Technology (HSST) Program is to develop and validate technology applicable to quantitative assessments of fracture prevention margins in nuclear reactor pressure vessels (RPVs) containing flaws and subjected to service-induced material toughness degradation. This paper describes an experimental/analytical program for the development of a Weibull statistical model of cleavage fracture toughness for applications to shallow surface-breaking and embedded flaws in RPV materials subjected to multi-axial loading conditions. The experimental part includes both material characterization testing and larger fracture toughness experiments conducted using a special-purpose cruciform beam specimen developed by Oak Ridge National Laboratory for applying biaxial loads to shallow cracks. Test materials (pressure vessel steels) included plate product forms (conforming to ASTM A533 Grade B Class 1 specifications) and shell segments procured from a pressurized-water reactor vessel intended for a nuclear power plant. Results from tests performed on cruciform specimens demonstrated that biaxial loading can have a pronounced effect on shallow-flaw fracture toughness in the lower-transition temperature region. A local approach methodology based on a three-parameter Weibull model was developed to correlate these experimentally-observed biaxial effects on fracture toughness. The Weibull model, combined with a new hydrostatic stress criterion in place of the more commonly used maximum principal stress in the kernel of the Weibull stress integral definition, is shown to provide a scaling mechanism between uniaxial and biaxial loading states for 2-dimensional flaws located in the A533-B plate material. The Weibull stress density was introduced as a matrice for identifying regions along a semi-elliptical flaw front that have a higher probability of cleavage initiation. Cumulative

Pressure vessel integrity 1991

International Nuclear Information System (INIS)

Bhandari, S.; Doney, R.O.; McDonald, M.S.; Jones, D.P.; Wilson, W.K.; Pennell, W.E.

1991-01-01

This volume contains papers relating to the structural integrity assessment of pressure vessels and piping, with special emphasis on nuclear industry applications. The papers were prepared for technical sessions developed under the sponsorship of the ASME Pressure Vessels and Piping Division Committees for Codes and Standards, Computer Technology, Design and Analysis, and Materials Fabrication. They were presented at the 1991 Pressure Vessels and Piping Division Conference in San Diego, California, June 23-27. The primary objective of the sponsoring organization is to provide a forum for the dissemination and discussion of information on development and application of technology for the structural integrity assessment of pressure vessels and piping. This publication includes contributions from authors from Australia, France, Japan, Sweden, Switzerland, the United Kingdom, and the United States. The papers here are organized in six sections, each with a particular emphasis as indicated in the following section titles: Fracture Technology Status and Application Experience; Crack Initiation, Propagation and Arrest; Ductile Tearing; Constraint, Stress State, and Local-Brittle-Zones Effects; Computational Techniques for Fracture and Corrosion Fatigue; and Codes and Standards for Fatigue, Fracture and Erosion/Corrosion

A simplified hydrodynamic model of hydrogen flame propagation in reactor vessels

International Nuclear Information System (INIS)

Baer, M.; Ratzel, A.

1983-01-01

A hydrodynamic model for hydrogen flame propagation in reactor geometries is presented. This model is consistent with the theory of slow combustion in which the gasdynamic field equations are treated in the limit of small Mach numbers. To the lowest order, pressure is spatially uniform. The flame is treated as a density and entropy discontinuity which propagates at prescribed burning velocities, corresponding to laminar or turbulent flames. Radiation cooling of the burned combustion gases and possible preheating of the unburned gases during propagation of the flame is included using a molecular gas-band thermal radiation model. Application of this model has been developed for 1-D variable area flame propagation. Multidimensional effects induced by hydrodynamics and buoyancy are introduced as a correction to the burn velocity (which reflects a modification of planar flame surface to a distorted surface) using experimentally measured pressure-rise time data for hydrogen/air deflagrations in cylindrical vessels. This semianalytical model of flame propagation reduces to a set of ordinary differential equations which describes the temporal variations of vessel pressure, burned volume and gas entropy. The thermodynamic state of the burned gas immediately following the flame is determined using an isobaric Hugoniot relationship. At other locations the burned gas thermodynamic states are determined using a Lagrangian particle tracking method. Results of a computer code using the method are presented

Reactor pressure vessel embrittlement: Insights from neural network modelling

Science.gov (United States)

Mathew, J.; Parfitt, D.; Wilford, K.; Riddle, N.; Alamaniotis, M.; Chroneos, A.; Fitzpatrick, M. E.

2018-04-01

Irradiation embrittlement of steel pressure vessels is an important consideration for the operation of current and future light water nuclear reactors. In this study we employ an ensemble of artificial neural networks in order to provide predictions of the embrittlement using two literature datasets, one based on US surveillance data and the second from the IVAR experiment. We use these networks to examine trends with input variables and to assess various literature models including compositional effects and the role of flux and temperature. Overall, the networks agree with the existing literature models and we comment on their more general use in predicting irradiation embrittlement.

Effects of no stiffness inside unbonded tendon ducts on the behavior of prestressd concrete containment vessels

Energy Technology Data Exchange (ETDEWEB)

Noh, Sang Hoon; Kwak, Hyo Gyong [Korea Advanced Institute of Science and Technology, Daejeon (Korea, Republic of); Jung, Rae Young; Noh, Sang Hoon [Korea Institute of Nuclear Safety, Daejeon (Korea, Republic of)

2016-06-15

The numerical simulation methodologies to evaluate the structural behaviors of prestressed concrete containment vessels (PCCVs) have been substantially developed in recent decades. However, there remain several issues to be investigated more closely to narrow the gap between test results and numerical simulations. As one of those issues, the effects of no stiffness inside unbonded tendon ducts on the behavior of PCCVs are investigated in this study. Duct holes for prestressing cables' passing are provided inside the containment wall and dome in one to three directions for general PCCVs. The specific stress distribution along the periphery of the prestressing duct hole and the loss of stiffness inside the hole, especially in an unbonded tendon system, are usually neglected in the analysis of PCCVs with the assumption that the duct hole is filled with concrete. However, duct holes are not small enough to be neglected. In this study, the effects of no stiffness inside the unbonded tendon system on the behaviors of PCCVs are evaluated using both analytical and numerical approaches. From the results, the effects of no stiffness in unbonded tendons need to be considered in numerical simulations for PCCVs, especially under internal pressure loading.

Lessons Learned Following the Successful Decommissioning of a Reaction Vessel Containing Lime Sludge and Technetium-99

International Nuclear Information System (INIS)

Dawson, P. M.; Watson, D. D.; Hylko, J. M.

2002-01-01

This paper documents how WESKEM, LLC utilized available source term information, integrated safety management, and associated project controls to safely decommission a reaction vessel and repackage sludge containing various Resource Conservation and Recovery Act constituents and technetium-99 (Tc-99). The decommissioning activities were segmented into five separate stages, allowing the project team to control work related decisions based on their knowledge, experience, expertise, and field observations. The information and experience gained from each previous stage and rehearsals contributed to modifying subsequent entries, further emphasizing the importance of developing hold points and incorporating lessons learned. The hold points and lessons learned, such as performing detailed personal protective equipment (PPE) inspections during sizing and repackaging operations, and using foam-type piping insulation to prevent workers from cutting or puncturing their PPE on sharp edge s or small shards generated during sizing operations, minimized direct contact with the Tc-99. To prevent the spread of contamination, the decommissioning activities were performed inside a containment enclosure connected to negative air machines. After performing over 235 individual entries totaling over 285 project hours, only one first aid was recorded during this five-stage project

Effects of no stiffness inside unbonded tendon ducts on the behavior of prestressd concrete containment vessels

International Nuclear Information System (INIS)

Noh, Sang Hoon; Kwak, Hyo Gyong; Jung, Rae Young; Noh, Sang Hoon

2016-01-01

The numerical simulation methodologies to evaluate the structural behaviors of prestressed concrete containment vessels (PCCVs) have been substantially developed in recent decades. However, there remain several issues to be investigated more closely to narrow the gap between test results and numerical simulations. As one of those issues, the effects of no stiffness inside unbonded tendon ducts on the behavior of PCCVs are investigated in this study. Duct holes for prestressing cables' passing are provided inside the containment wall and dome in one to three directions for general PCCVs. The specific stress distribution along the periphery of the prestressing duct hole and the loss of stiffness inside the hole, especially in an unbonded tendon system, are usually neglected in the analysis of PCCVs with the assumption that the duct hole is filled with concrete. However, duct holes are not small enough to be neglected. In this study, the effects of no stiffness inside the unbonded tendon system on the behaviors of PCCVs are evaluated using both analytical and numerical approaches. From the results, the effects of no stiffness in unbonded tendons need to be considered in numerical simulations for PCCVs, especially under internal pressure loading

Conjugate heat transfer analysis for in-vessel retention with external reactor vessel cooling

International Nuclear Information System (INIS)

Park, Jong-Woon; Bae, Jae-ho; Song, Hyuk-Jin

2016-01-01

Highlights: • A conjugate heat transfer analysis method is applied for in-vessel corium retention. • 3D heat diffusion has a formidable effect in alleviating focusing heat load from metallic layer. • The focusing heat load is decreased by about 2.5 times on the external surface. - Abstract: A conjugate heat transfer analysis method for the thermal integrity of a reactor vessel under external reactor vessel cooling conditions is developed to resolve light metal layer focusing effect issue for in-vessel retention. The method calculates steady-state three-dimensional temperature distribution of a reactor vessel using coupled conjugate heat transfer between in-vessel three-layered stratified corium (metallic pool, oxide pool and heavy metal and polar-angle dependent boiling heat transfer at the outer surface of a reactor vessel). The three-layer corium heat transfer model is utilizing lumped-parameter thermal-resistance circuit method. For the ex-vessel boiling boundary conditions, nucleate, transition and film boiling are considered. The thermal integrity of a reactor vessel is addressed in terms of heat flux at the outer-most nodes of the vessel and remaining thickness profile. The vessel three-dimensional heat conduction is validated against a commercial code. It is found that even though the internal heat flux from the metal layer goes far beyond critical heat flux (CHF) the heat flux from the outermost nodes of the vessel may be maintained below CHF due to massive vessel heat diffusion. The heat diffusion throughout the vessel is more pronounced for relatively low heat generation rate in an oxide pool. Parametric calculations are performed considering thermal conditions such as peak heat flux from a light metal layer, heat generation in an oxide pool and external boiling conditions. The major finding is that the most crucial factor for success of in-vessel retention is not the mass of the molten light metal above the oxide pool but the heat generation rate

Commercial Passenger Fishing Vessel Fishery

Data.gov (United States)

National Oceanic and Atmospheric Administration, Department of Commerce — This data set contains the logbook data from U.S.A. Commercial Passenger Fishing Vessels (CPFV) fishing in the U.S.A. EEZ and in waters off of Baja California, from...

The analysis of reactor vessel surveillance program data

International Nuclear Information System (INIS)

Norris, E.B.

1979-01-01

Commercial nuclear power reactor vessel surveillance programs are provided by the reactor supplier and are designed to meet the requirements of ASTM Method E 185. (3). Each surveillance capsule contains sets of Charpy V-notch (Csub(v)) specimens representing selected materials from the vessel beltline region and some reference steel, tension test specimens machined from selected beltline materials, temperature monitors, and neutron flux dosimeters. Surveillance capsules may also contain fracture mechanics specimens machined from selected vessel beltline materials. The major steps in the conduct of a surveillance program include (1) the testing of the surveillance specimens to determine the exposure conditions at the capsule location and the resulting embrittlement of the vessel steel, (2) the extrapolation of the capsule results to the pressure vessel wall, and (3) the determination of the heatup and cooldown limits for normal, upset, and test operation. This paper will present data obtained from commercial light water reactor surveillance programs to illustrate the methods of analysis currently in use at Southwest Research Institute and to demonstrate some of the limitations imposed by the data available. Details concerning the procedures for testing the surveillance capsule specimens will not be included because they are considered to be outside of the scope of this paper

The development of evaporative liquid film model for analysis of passive containment cooling system

Energy Technology Data Exchange (ETDEWEB)

Park, Hong June; Hwang, Young Dong; Kim, Hee Cheol; Kim, Young In; Chang, Moon Hee

2000-07-01

An analytical model was developed to simulate behavior of the liquid film formed on the outside surface of the steel containment vessel of PCCS including the ellipsoidal dome and the vertical wall. The model was coupled with CFX code using the user subroutines provided by the code, and a series of numerical calculations were performed to evaluate the evaporative heat transfer coefficient at the interface. Numerical results for Sherwood number and evaporative heat transfer coefficient were compared with the experimental data. The results were in good agreement with the experimental data. The calculated liquid film thickness showed good agreement with that of Sun except an upper portion of the channel. The model was applied to the full scale of PCCS to investigate the effects of dome and chimney on the evaporation rate. The results showed that the heat transfer coefficient in the dome region, where the flow cross-sectional area decreases and the swirling occurs, was lower than that of the vertical annulus region. The calculated evaporative heat transfer coefficient was about 20 times larger than that of the dry cooling. Sensitivity studies on the gap size and the wall temperature were also performed to figure out their effects on the heat transfer coefficient and inlet air average velocity. Through the analysis of the dryout point, the minimum liquid film flow rate to cover the entire surface of the vessel was estimated.

The development of evaporative liquid film model for analysis of passive containment cooling system

International Nuclear Information System (INIS)

Park, Hong June; Hwang, Young Dong; Kim, Hee Cheol; Kim, Young In; Chang, Moon Hee

2000-07-01

An analytical model was developed to simulate behavior of the liquid film formed on the outside surface of the steel containment vessel of PCCS including the ellipsoidal dome and the vertical wall. The model was coupled with CFX code using the user subroutines provided by the code, and a series of numerical calculations were performed to evaluate the evaporative heat transfer coefficient at the interface. Numerical results for Sherwood number and evaporative heat transfer coefficient were compared with the experimental data. The results were in good agreement with the experimental data. The calculated liquid film thickness showed good agreement with that of Sun except an upper portion of the channel. The model was applied to the full scale of PCCS to investigate the effects of dome and chimney on the evaporation rate. The results showed that the heat transfer coefficient in the dome region, where the flow cross-sectional area decreases and the swirling occurs, was lower than that of the vertical annulus region. The calculated evaporative heat transfer coefficient was about 20 times larger than that of the dry cooling. Sensitivity studies on the gap size and the wall temperature were also performed to figure out their effects on the heat transfer coefficient and inlet air average velocity. Through the analysis of the dryout point, the minimum liquid film flow rate to cover the entire surface of the vessel was estimated

CFD Modeling of a Multiphase Gravity Separator Vessel

KAUST Repository

Narayan, Gautham

2017-05-23

The poster highlights a CFD study that incorporates a combined Eulerian multi-fluid multiphase and a Population Balance Model (PBM) to study the flow inside a typical multiphase gravity separator vessel (GSV) found in oil and gas industry. The simulations were performed using Ansys Fluent CFD package running on KAUST supercomputer, Shaheen. Also, a highlight of a scalability study is presented. The effect of I/O bottlenecks and using Hierarchical Data Format (HDF5) for collective and independent parallel reading of case file is presented. This work is an outcome of a research collaboration on an Aramco project on Shaheen.

CFD Modeling of a Multiphase Gravity Separator Vessel

KAUST Repository

Narayan, Gautham; Khurram, Rooh Ul Amin; Elsaadawy, Ehab

2017-01-01

The poster highlights a CFD study that incorporates a combined Eulerian multi-fluid multiphase and a Population Balance Model (PBM) to study the flow inside a typical multiphase gravity separator vessel (GSV) found in oil and gas industry. The simulations were performed using Ansys Fluent CFD package running on KAUST supercomputer, Shaheen. Also, a highlight of a scalability study is presented. The effect of I/O bottlenecks and using Hierarchical Data Format (HDF5) for collective and independent parallel reading of case file is presented. This work is an outcome of a research collaboration on an Aramco project on Shaheen.

Guidelines for pressure vessel safety assessment

Science.gov (United States)

Yukawa, S.

1990-04-01

A technical overview and information on metallic pressure containment vessels and tanks is given. The intent is to provide Occupational Safety and Health Administration (OSHA) personnel and other persons with information to assist in the evaluation of the safety of operating pressure vessels and low pressure storage tanks. The scope is limited to general industrial application vessels and tanks constructed of carbon or low alloy steels and used at temperatures between -75 and 315 C (-100 and 600 F). Information on design codes, materials, fabrication processes, inspection and testing applicable to the vessels and tanks are presented. The majority of the vessels and tanks are made to the rules and requirements of ASME Code Section VIII or API Standard 620. The causes of deterioration and damage in operation are described and methods and capabilities of detecting serious damage and cracking are discussed. Guidelines and recommendations formulated by various groups to inspect for the damages being found and to mitigate the causes and effects of the problems are presented.

Nuclear reactor installation with outer shell enclosing a primary pressure vessel

International Nuclear Information System (INIS)

1975-01-01

The high temperature nuclear reactor installation described includes a fluid cooled nuclear heat source, a primary pressure vessel and outer shell around the primary pressure vessel and acting as a protection for it against outside projectiles. A floor is provided internally dividing the outside shell into two upper and lower sections and an inside wall dividing the lower section into one part containing the primary pressure vessel and a second part, both made pressure tight with respect to each other and with the outside shell and forming with the latter a secondary means of containment [fr

49 CFR 176.76 - Transport vehicles, freight containers, and portable tanks containing hazardous materials.

Science.gov (United States)

2010-10-01

... 49 Transportation 2 2010-10-01 2010-10-01 false Transport vehicles, freight containers, and... TRANSPORTATION HAZARDOUS MATERIALS REGULATIONS CARRIAGE BY VESSEL General Handling and Stowage § 176.76 Transport... paragraphs (b) through (f) of this section, hazardous materials authorized to be transported by vessel may be...

Study on prediction model of irradiation embrittlement for reactor pressure vessel steel

International Nuclear Information System (INIS)

Wang Rongshan; Xu Chaoliang; Huang Ping; Liu Xiangbing; Ren Ai; Chen Jun; Li Chengliang

2014-01-01

The study on prediction model of irradiation embrittlement for reactor pres- sure vessel (RPV) steel is an important method for long term operation. According to the deep analysis of the previous prediction models developed worldwide, the drawbacks of these models were given and a new irradiation embrittlement prediction model PMIE-2012 was developed. A corresponding reliability assessment was carried out by irradiation surveillance data. The assessment results show that the PMIE-2012 have a high reliability and accuracy on irradiation embrittlement prediction. (authors)

Underwater Shock Response Analysis of a Floating Vessel

Directory of Open Access Journals (Sweden)

J.E. van Aanhold

1998-01-01

Full Text Available The response of a surface vessel to underwater shock has been calculated using an explicit finite element analysis. The analysis model is two-dimensional and contains the floating steel structure, a large surrounding water volume and the free surface. The underwater shock is applied in the form of a plane shock wave and cavitation is considered in the analysis. Advanced computer graphics, in particular video animations, provide a powerful and indispensable means for the presentation and evaluation of the analysis results.

Numerical analysis of coolant mixing in the pressure vessel of WWER-440 type nuclear reactors

International Nuclear Information System (INIS)

Boros, I.; Aszodi, A.

2003-01-01

The precise description of the coolant mixing processes taking place in the reactor pressure vessel (RPV) of pressurized water nuclear reactors has an essential importance during power operation, as well as in case of incidental or accidental conditions. In this paper the detailed CFD model of the pressure vessel of a WWER-440 type reactor and calculations performed with this RPV model are presented. The CFD model of the pressure vessel contains all the important internal structural elements of the RPV. Sensitivity study on the effect of these elements was also carried out. Both steady-state and transient calculation were performed using the CFD code CFX-5.5.1. The results of the steady-state calculations give the so called mixing factors, i.e. the effect of each single primary loop at the core inlet. The mixing factors can be given for nominal circumstances (i.e. all main coolant pumps are working) or in case of less than six working MCPs. In order to validate the model the calculated mixing factors are compared with the values measured in the Paks NPP (Authors)

Draft paper: On the analysis of diffusive mass transfer in ex-vessel corium pools

International Nuclear Information System (INIS)

Frolov, Kyrill N.

2003-01-01

In case of a severe accident at a nuclear power plant (NPP) involving the reactor pressure vessel (RPV) melt-through, confident solidification of ex-vessel corium is the imperative condition of its safe retention within the plant containment. The rate-determining process for solidification of ex-vessel coriums in the long-term is the chemical diffusion in the liquid phase at the solid-liquid interface. The process of chemical diffusion in the diffusive boundary layer can evolve taking on different rates, depending on the boundary conditions and the melt composition. Nonetheless, the chemical diffusion rates would entwine the self-diffusivities of corium constituents, which in turn would depend on the melt chemical composition. This work looks at some aspects of analytical and experimental determination of self-diffusivities of corium constituents. Following the corium-concrete interaction, an ex-vessel corium melt would contain several chemical components, including a fraction of silica. Accordingly, ex-vessel corium is considered in this paper as a silicate melts. In the realm of the geological and glass sciences, where silicate melts are most often discussed, the diffusive transport and viscous flow are conceived interrelated from a phenomenological point of view. Though the viscous and diffusive mass transfer mechanisms are not identical for different species even in the same melt, a combination of semi-empirical models can still provide an estimation of the diffusion thresholds in ex-vessel corium melts. Thus, the first part of this paper presents an analysis of the applicability of such empirical models for simple silicate melts based on the published data. This is followed by an estimation of diffusivities in melt compositions typical of ex-vessel coriums. Alternatively, although the general trend towards a coupled description of the viscous flow and diffusion for ex-vessel corium melts seems promising, it is limited to published data on self-diffusivities of

Discharge of Non-Reactive Fluids from Vessels

Directory of Open Access Journals (Sweden)

M. Castier

Full Text Available Abstract This paper presents simulations of discharges from pressure vessels that consistently account for non-ideal fluid behavior in all the required thermodynamic properties and individually considers all the chemical components present. The underlying assumption is that phase equilibrium occurs instantaneously inside the vessel and, thus, the dynamics of the fluid in the vessel comprises a sequence of equilibrium states. The formulation leads to a system of differential-algebraic equations in which the component mass balances and the energy balance are ordinary differential equations. The algebraic equations account for the phase equilibrium conditions inside the vessel and at the discharge point, and for sound speed calculations. The simulator allows detailed predictions of the condition inside the vessel and at the discharge point as a function of time, including the flow rate and composition of the discharge. The paper presents conceptual applications of the simulator to predict the effect of leaks from vessels containing mixtures of light gases and/or hydrocarbons and comparisons to experimental data.

Loads on EPR containment after RPV failure at high pressure

International Nuclear Information System (INIS)

Jacobs, G.

1995-01-01

As regards the desgin of the EPR, the general strategy is to eliminate, the vessel failure at high pressure by preventive and mitigative measures. The design proposals involved trust in the reliability of dedicated devices (relief valves) for rapid depressurization. The aim is to attain a lower pressure level at the moment of vessel failure, so that the containment is capable to cope with the blowdown impact on the pit walls and the vessel supporting structures. Nevertheless, the potential of a high-pressure failure of the vessel must be kept in mind, whatever well thought-out and reliable preventive depressurization measures might be. Therefore, the reactor pressure blowdown has been studied in order to quantify the ultimate containment load, which might support future design requirements. The calculations were performed with the LWR transient analysis thermal-hydraulics computer code REALAP5/MOD3. In previous analyses, the nodalization of the problem was based on the geometrical conditions of a typical German 1300 MW(e) NPP. In the present analysis a new input model has been used, which was based on the EPR conditions. (orig./HP)

Cholinergic innervation of human mesenteric lymphatic vessels.

Science.gov (United States)

D'Andrea, V; Bianchi, E; Taurone, S; Mignini, F; Cavallotti, C; Artico, M

2013-11-01

The cholinergic neurotransmission within the human mesenteric lymphatic vessels has been poorly studied. Therefore, our aim is to analyse the cholinergic nerve fibres of lymphatic vessels using the traditional enzymatic techniques of staining, plus the biochemical modifications of acetylcholinesterase (AChE) activity. Specimens obtained from human mesenteric lymphatic vessels were subjected to the following experimental procedures: 1) drawing, cutting and staining of tissues; 2) staining of total nerve fibres; 3) enzymatic staining of cholinergic nerve fibres; 4) homogenisation of tissues; 5) biochemical amount of proteins; 6) biochemical amount of AChE activity; 6) quantitative analysis of images; 7) statistical analysis of data. The mesenteric lymphatic vessels show many AChE positive nerve fibres around their wall with an almost plexiform distribution. The incubation time was performed at 1 h (partial activity) and 6 h (total activity). Moreover, biochemical dosage of the same enzymatic activity confirms the results obtained with morphological methods. The homogenates of the studied tissues contain strong AChE activity. In our study, the lymphatic vessels appeared to contain few cholinergic nerve fibres. Therefore, it is expected that perivascular nerve stimulation stimulates cholinergic nerves innervating the mesenteric arteries to release the neurotransmitter AChE, which activates muscarinic or nicotinic receptors to modulate adrenergic neurotransmission. These results strongly suggest, that perivascular cholinergic nerves have little or no effect on the adrenergic nerve function in mesenteric arteries. The cholinergic nerves innervating mesenteric arteries do not mediate direct vascular responses.

Earthquake-proof and anti-vibration device for nuclear containers

International Nuclear Information System (INIS)

Hayano, Mutsuhiko; Imayoshi, Sho; Hirota, Koichi.

1985-01-01

Purpose: To surely support the reactor vessel by a lower support structure upon earthquakes. Constitution: Low melting alloys are filled to a gap between the lower ring of a reactor container and a retainer and to a gap at the vertical portion between the protrusion of a guard vessel and the lower support structure. The liquid level of the low melting metal in the guard vessel upon melting is set such that it forms a liquid level filling the gap between the lower ring of the reactor container and the retainer even if the reactor vessel is axially shrinked or expanded. By filling the low melting metal in this way into the gap, the temperature-dependency of the defined members is improved and the support for the vessel or container upon earthquake can be made reliable due to the structural material. (Yoshino, Y.)

Analytical model for shear strength of end slabs of prestressed concrete nuclear reactor vessels

International Nuclear Information System (INIS)

Abdulrahman, H.O.; Sozen, M.A.; Schnobrich, W.C.

1979-04-01

The results are presented of an investigation of the behavior and strength of flat end slabs of cylindrical prestressed concrete nuclear reactor vessels. The investigation included tests of ten small-scale pressure vessels and development of a nonlinear finite-element model to simulate the deformation response and strength of the end slabs. Because earlier experimental studies had shown that the flexural strength of the end slab could be calculated using intelligible procedures, the emphasis of this investigation was on shear strength

Venting device for nuclear reactor container

International Nuclear Information System (INIS)

Yamashita, Masahiro; Ogata, Ken-ichi.

1994-01-01

An airtight vessel of a venting device of a nuclear reactor container is connected with a reactor container by way of a communication pipeline. A feed water tank is disposed at a position higher than the liquid surface of scrubbing water in the airtight vessel for supplying scrubbing water to the airtight vessel. In addition, a scrubbing water storage tank is disposed at a position hither than the feed water tank for supplying scrubbing water to the feed water tank. Storage water in the feed water tank is introduced into the airtight vessel by the predetermined opening operation of a valve by the pressure exerted on the liquid surface and the own weight of the storage water. Further, the storage water in the scrubbing water storage tank is led into the feed water tank by the water head pressure. The scrubbing water for keeping the performance of the venting device of the reactor container can be supplied by a highly reliable method without using AC power source or the like as a driving source. (I.N.)

Passive containment cooling water distribution device

Science.gov (United States)

Conway, Lawrence E.; Fanto, Susan V.

1994-01-01

A passive containment cooling system for a nuclear reactor containment vessel. Disclosed is a cooling water distribution system for introducing cooling water by gravity uniformly over the outer surface of a steel containment vessel using a series of radial guide elements and cascading weir boxes to collect and then distribute the cooling water into a series of distribution areas through a plurality of cascading weirs. The cooling water is then uniformly distributed over the curved surface by a plurality of weir notches in the face plate of the weir box.

A mathematical model for cost of maritime transport. Application to competitiveness of nuclear vessels

International Nuclear Information System (INIS)

Dorval, C.

1966-05-01

In studying the competitiveness of a nuclear merchant vessel, economic assessments in terms of figures were discarded in favor of a simplified model, which gives a clearer idea of the mechanism of the comparison between alternative vessels and the particular influence of each parameter. An expression is formulated for the unit cost per ton carried over a given distance as a function of the variables (speed and deadweight tonnage) and is used to determine the optima for conventional and nuclear vessels. To represent the freight market involved in the optimization studies, and thus in the competitiveness computation, two cases are taken into account: the tonnage to be carried annually is limited, and the tonnage to be carried annually is not limited. In both cases the optima are calculated and compared for a conventional and a nuclear vessel. Competitiveness curves are plotted as a function of the ratios of nuclear and conventional fuel costs and nuclear and conventional marginal power costs. These curves express the limiting values of the above two ratios for which the transport costs of the nuclear and conventional vessels are equal. The competitiveness curves vary considerably according to the hypothesis adopted for the freight market and the limit of tonnage carried annually. (author) [fr

Nuclear power plant pressure vessels. Control of piping

International Nuclear Information System (INIS)

2000-01-01

The guide presents requirements for the pipework of nuclear facilities in Finland. According to the section 117 of the Finnish Nuclear Energy Degree (161/88), the Radiation and Nuclear Safety Authority of Finland (STUK) controls the pressure vessels of nuclear facilities in accordance with the Nuclear Energy Act (990/87) and, to the extent applicable in accordance with the Act of Pressure Vessels (98/73) and the rules and regulations issued by the virtue of these. In addition STUK is an inspecting authority of pressure vessels of nuclear facilities in accordance with the Pressure Vessel Degree (549/1973). According to the section of the Pressure Vessel Degree, a pressure vessel is a steam boiler, pressure container, pipework of other such appliance in which the pressure is above or may come to exceed the atmospheric pressure. Guide YVL 3.0 describes in general terms how STUK controls pressure vessels. STUK controls Safety Class 1, 2 and 3 piping as well as Class EYT (non-nuclear) and their support structures in accordance with this guide and applies the provisions of the Decision of the Ministry of Trade and Industry on piping (71/1975) issued by virtue of the Pressure Vessel Decree

Segmentation of vessels cluttered with cells using a physics based model.

Science.gov (United States)

Schmugge, Stephen J; Keller, Steve; Nguyen, Nhat; Souvenir, Richard; Huynh, Toan; Clemens, Mark; Shin, Min C

2008-01-01

Segmentation of vessels in biomedical images is important as it can provide insight into analysis of vascular morphology, topology and is required for kinetic analysis of flow velocity and vessel permeability. Intravital microscopy is a powerful tool as it enables in vivo imaging of both vasculature and circulating cells. However, the analysis of vasculature in those images is difficult due to the presence of cells and their image gradient. In this paper, we provide a novel method of segmenting vessels with a high level of cell related clutter. A set of virtual point pairs ("vessel probes") are moved reacting to forces including Vessel Vector Flow (VVF) and Vessel Boundary Vector Flow (VBVF) forces. Incorporating the cell detection, the VVF force attracts the probes toward the vessel, while the VBVF force attracts the virtual points of the probes to localize the vessel boundary without being distracted by the image features of the cells. The vessel probes are moved according to Newtonian Physics reacting to the net of forces applied on them. We demonstrate the results on a set of five real in vivo images of liver vasculature cluttered by white blood cells. When compared against the ground truth prepared by the technician, the Root Mean Squared Error (RMSE) of segmentation with VVF and VBVF was 55% lower than the method without VVF and VBVF.

In-vessel core debris retention through external flooding of the reactor pressure vessel. State-of-the-art report

Energy Technology Data Exchange (ETDEWEB)

Heel, A.M.J.M. van

1995-07-01

An overview of the state-of-the-art knowledge on the ex-vessel flooding accident management strategy for severe accidents in a NPP has been given. The feasibility has been discussed, as well as the in- and ex-vessel phenomena, which influence the structural integrity of the vessel. Finally, some computer codes with the ability to model the phenomena involved in ex-vessel flooding have been discussed. (orig./HP).

In-vessel core debris retention through external flooding of the reactor pressure vessel. State-of-the-art report

International Nuclear Information System (INIS)

Heel, A.M.J.M. van.

1995-07-01

An overview of the state-of-the-art knowledge on the ex-vessel flooding accident management strategy for severe accidents in a NPP has been given. The feasibility has been discussed, as well as the in- and ex-vessel phenomena, which influence the structural integrity of the vessel. Finally, some computer codes with the ability to model the phenomena involved in ex-vessel flooding have been discussed. (orig./HP)

Assessment of models predicting irradiation effects on tensile properties of reactor pressure vessel steels

International Nuclear Information System (INIS)

Pineau, L.; Landron, C.

2015-01-01

In this paper, an analysis of tensile data acquired as part of the French Reactor Vessel Surveillance Program (RVSP) is produced. This program contains amongst other mechanical tests, tensile tests at 20 and 300 C degrees on non irradiated base metals and at 300 C degrees only on irradiated materials. It shows that irradiation leads to an increase in the yield strength and a decrease in the strain hardening. The exploitation of tensile results has permitted to express a relationship between yield strength increase measured and fluence value, as well as between strain hardening decrease and yield strength evolution. The use of these relations in the aim at predicting evolution of tensile properties with irradiation has then permitted to propose a methodology to model entire stress-strain curves of irradiated base metal only based on the non irradiated stress-strain curve. These predictions were successfully compared with an experimental standard case. (authors)

Analyses and testing of model prestressed concrete reactor vessels with built-in planes of weakness

International Nuclear Information System (INIS)

Dawson, P.; Paton, A.A.; Fleischer, C.C.

1990-01-01

This paper describes the design, construction, analyses and testing of two small scale, single cavity prestressed concrete reactor vessel models, one without planes of weakness and one with planes of weakness immediately behind the cavity liner. This work was carried out to extend a previous study which had suggested the likely feasibility of constructing regions of prestressed concrete reactor vessels and biological shields, which become activated, using easily removable blocks, separated by a suitable membrane. The paper describes the results obtained and concludes that the planes of weakness concept could offer a means of facilitating the dismantling of activated regions of prestressed concrete reactor vessels, biological shields and similar types of structure. (author)

Large Pelagic Logbook Set Survey (Vessels)

Data.gov (United States)

National Oceanic and Atmospheric Administration, Department of Commerce — This data set contains catch and effort for fishing trips that are taken by vessels with a Federal permit issued for the swordfish and sharks under the Highly...

Large Pelagic Logbook Trip Survey (Vessels)

Data.gov (United States)

National Oceanic and Atmospheric Administration, Department of Commerce — This data set contains catch and effort for fishing trips that are taken by vessels with a Federal permit issued for the swordfish and sharks under the Highly...

Hydrogen combustion study in the containment of Atucha-I nuclear power plant

International Nuclear Information System (INIS)

Baron, J.H.; Gonzalez Videla, E.

1997-01-01

In this paper the combustion of hydrogen was modeled and studied in the containment vessel of the Atucha I nuclear power station using the CONTAIN package. The hydrogen comes from the oxidation of metallic materials during the severe accidents proposed. The CONTAIN package is an integrated tool that analyzes the physical, chemical and radiation conditions that affect the containment structure of the radioactive materials unloaded from the primary system during a severe accident in the reactor. (author) [es

Reactor container cooling device

Energy Technology Data Exchange (ETDEWEB)

Ando, Koji; Kinoshita, Shoichiro

1995-11-10

The device of the present invention efficiently lowers pressure and temperature in a reactor container upon occurrence of a severe accident in a BWR-type reactor and can cool the inside of the container for a long period of time. That is, (1) pipelines on the side of an exhaustion tower of a filter portion in a filter bent device of the reactor container are in communication with pipelines on the side of a steam inlet of a static container cooling device by way of horizontal pipelines, (2) a back flow check valve is disposed to horizontal pipelines, (3) a steam discharge valve for a pressure vessel is disposed closer to the reactor container than the joint portion between the pipelines on the side of the steam inlet and the horizontal pipelines. Upon occurrence of a severe accident, when the pressure vessel should be ruptured and steams containing aerosol in the reactor core should be filled in the reactor container, the inlet valve of the static container cooling device is closed. Steams are flown into the filter bent device of the reactor container, where the aerosols can be removed. (I.S.).

Component nuclear containment structure

International Nuclear Information System (INIS)

Harstead, G.A.

1979-01-01

The invention described is intended for use primarily as a nuclear containment structure. Such structures are required to surround the nuclear steam supply system and to contain the effects of breaks in the nuclear steam supply system, or i.e. loss of coolant accidents. Nuclear containment structures are required to withstand internal pressure and temperatures which result from loss of coolant accidents, and to provide for radiation shielding during operation and during the loss of coolant accident, as well as to resist all other applied loads, such as earthquakes. The nuclear containment structure described herein is a composite nuclear containment structure, and is one which structurally combines two previous systems; namely, a steel vessel, and a lined concrete structure. The steel vessel provides strength to resist internal pressure and accommodate temperature increases, the lined concrete structure provides resistance to internal pressure by having a liner which will prevent leakage, and which is in contact with the concrete structure which provides the strength to resist the pressure

Modulation of lung inflammation by vessel dilator in a mouse model of allergic asthma

Directory of Open Access Journals (Sweden)

Cormier Stephania A

2009-07-01

Full Text Available Abstract Background Atrial natriuretic peptide (ANP and its receptor, NPRA, have been extensively studied in terms of cardiovascular effects. We have found that the ANP-NPRA signaling pathway is also involved in airway allergic inflammation and asthma. ANP, a C-terminal peptide (amino acid 99–126 of pro-atrial natriuretic factor (proANF and a recombinant peptide, NP73-102 (amino acid 73–102 of proANF have been reported to induce bronchoprotective effects in a mouse model of allergic asthma. In this report, we evaluated the effects of vessel dilator (VD, another N-terminal natriuretic peptide covering amino acids 31–67 of proANF, on acute lung inflammation in a mouse model of allergic asthma. Methods A549 cells were transfected with pVD or the pVAX1 control plasmid and cells were collected 24 hrs after transfection to analyze the effect of VD on inactivation of the extracellular-signal regulated receptor kinase (ERK1/2 through western blot. Luciferase assay, western blot and RT-PCR were also performed to analyze the effect of VD on NPRA expression. For determination of VD's attenuation of lung inflammation, BALB/c mice were sensitized and challenged with ovalbumin and then treated intranasally with chitosan nanoparticles containing pVD. Parameters of airway inflammation, such as airway hyperreactivity, proinflammatory cytokine levels, eosinophil recruitment and lung histopathology were compared with control mice receiving nanoparticles containing pVAX1 control plasmid. Results pVD nanoparticles inactivated ERK1/2 and downregulated NPRA expression in vitro, and intranasal treatment with pVD nanoparticles protected mice from airway inflammation. Conclusion VD's modulation of airway inflammation may result from its inactivation of ERK1/2 and downregulation of NPRA expression. Chitosan nanoparticles containing pVD may be therapeutically effective in preventing allergic airway inflammation.

Insights for aging management of light water reactor components: Metal containments

International Nuclear Information System (INIS)

Shah, V.N.; Sinha, U.P.; Smith, S.K.

1994-03-01

This report evaluates the available technical information and field experience related to management of aging damage to light water reactor metal containments. A generic aging management approach is suggested for the effective and comprehensive aging management of metal containments to ensure their safe operation. The major concern is corrosion of the embedded portion of the containment vessel and detection of this damage. The electromagnetic acoustic transducer and half-cell potential measurement are potential techniques to detect corrosion damage in the embedded portion of the containment vessel. Other corrosion-related concerns include inspection of corrosion damage on the inaccessible side of BWR Mark I and Mark II containment vessels and corrosion of the BWR Mark I torus and emergency core cooling system piping that penetrates the torus, and transgranular stress corrosion cracking of the penetration bellows. Fatigue-related concerns include reduction in the fatigue life (a) of a vessel caused by roughness of the corroded vessel surface and (b) of bellows because of any physical damage. Maintenance of surface coatings and sealant at the metal-concrete interface is the best protection against corrosion of the vessel

Thermal annealing of an embrittled reactor pressure vessel

International Nuclear Information System (INIS)

Mager, T.R.; Dragunov, Y.G.; Leitz, C.

1998-01-01

As a result of the popularity of the Agencies report 'Neutron Irradiation Embrittlement of Reactor Pressure Vessel Steels' of 1975, it was decided that another report on this broad subject would be of use. In this report, background and contemporary views on specially identified areas of the subject are considered as self-contained chapters, written by experts. Chapter 11 deals with thermal annealing of an embrittled reactor pressure vessel. Anneal procedures for vessels from both the US and the former USSR are mentioned schematically, wet anneals at lower temperature and dry anneals above RPV design temperatures are investigated. It is shown that heat treatment is a means of recovering mechanical properties which were degraded by neutron radiation exposure, thus assuring reactor pressure vessel compliance with regulatory requirements

Development and operational experiences of an automated remote inspection system for interior of primary containment vessel of a BWR

International Nuclear Information System (INIS)

Ozaki, N.; Chikara, S.; Fumio, T.; Katsuhiro, M.; Katsutoshi, S.; Ken-Ichiro, S.; Masaaki, F.; Masayoshi, S.

1983-01-01

A prototype was developed for an automated remote inspection system featuring continuous monitoring of the working status of major components inside the primary containment vessel of a boiling water reactor. This inspection system consists of four units, or vehicles, which are towed by a trolley chain along a monorail; a complex coaxial cable for data transmission and for power supply; and an operator's console. A TV camera, microphone, thermometer, hygrometer, and ionization chamber are mounted on the various units. After several months' testing under high-ambient temperature, the system was installed in the Tokai-2 power station of Japan Atomic Power Company for in situ tests

Non-linear dynamic response of reactor containment