Reactor container

International Nuclear Information System (INIS)

Shibata, Satoru; Kawashima, Hiroaki

1984-01-01

Purpose: To optimize the temperature distribution of the reactor container so as to moderate the thermal stress distribution on the reactor wall of LMFBR type reactor. Constitution: A good heat conductor (made of Al or Cu) is appended on the outer side of the reactor container wall from below the liquid level to the lower face of a deck plate. Further, heat insulators are disposed to the outside of the good heat conductor. Furthermore, a gas-cooling duct is circumferentially disposed at the contact portion between the good heat conductor and the deck plate around the reactor container. This enables to flow the cold heat from the liquid metal rapidly through the good heat conductor to the cooling duct and allows to maintain the temperature distribution on the reactor wall substantially linear even with the abrupt temperature change in the liquid metal. Further, by appending the good heat conductor covered with inactive metals not only on the outer side but also on the inside of the reactor wall to introduce the heat near the liquid level to the upper portion and escape the same to the cooling layer below the roof slab, the effect can be improved further. (Ikeda, J.)

Collecting and recirculating condensate in a nuclear reactor containment

International Nuclear Information System (INIS)

Schultz, T.L.

1993-01-01

An arrangement passively cools a nuclear reactor in the event of an emergency, condensing and recycling vaporized cooling water. The reactor is surrounded by a containment structure and has a storage tank for cooling liquid, such as water, vented to the containment structure by a port. The storage tank preferably is located inside the containment structure and is thermally coupleable to the reactor, e.g. by a heat exchanger, such that water in the storage tank is boiled off to carry away heat energy. The water is released as a vapor (steam) and condenses on the cooler interior surfaces of the containment structure. The condensed water flows downwardly due to gravity and is collected and routed back to the storage tank. One or more gutters are disposed along the interior wall of the containment structure for collecting the condensate from the wall. Piping is provided for communicating the condensate from the gutters to the storage tank. 3 figures

Collecting and recirculating condensate in a nuclear reactor containment

Science.gov (United States)

Schultz, Terry L.

1993-01-01

An arrangement passively cools a nuclear reactor in the event of an emergency, condensing and recycling vaporized cooling water. The reactor is surrounded by a containment structure and has a storage tank for cooling liquid, such as water, vented to the containment structure by a port. The storage tank preferably is located inside the containment structure and is thermally coupleable to the reactor, e.g. by a heat exchanger, such that water in the storage tank is boiled off to carry away heat energy. The water is released as a vapor (steam) and condenses on the cooler interior surfaces of the containment structure. The condensed water flows downwardly due to gravity and is collected and routed back to the storage tank. One or more gutters are disposed along the interior wall of the containment structure for collecting the condensate from the wall. Piping is provided for communicating the condensate from the gutters to the storage tank.

Specification for carbon and low alloy steel containment structures for stationary nuclear power reactors. [Now obsolescent (by Amendment No. 1)

Energy Technology Data Exchange (ETDEWEB)

1967-01-01

This British Standard covers the design, construction, inspection and testing of steel reactor containment structures made of carbon and low alloy steel for temperatures not exceeding 300 deg C. Such structures are not in contact with the reactor coolant during normal operation. Pressure-relieved structures are not excluded, provided they are of a form that contains the fission products or ensures their safe disposal. Attachments such as air-locks or piping that is or may become directly connected between the interior of the containment structure and a closure, and may therefore contain radioactive material released during accidents, is considered part of the containment structure.

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.).

Nuclear reactor container

International Nuclear Information System (INIS)

Ishiyama, Takenori.

1989-01-01

This invention concerns a nuclear reactor container in which heat is removed from a container by external water injection. Heat is removed from the container by immersing the lower portion of the container into water and scattering spary water from above. Thus, the container can be cooled by the spray water falling down along the outer wall of the container to condensate and cool vapors filled in the container upon occurrence of accidents. Further, since the inside of the container can be cooled also during usual operation, it can also serve as a dry well cooler. Accordingly, heat is removed from the reactor container upon occurrence of accidents by the automatic operation of a spray device corresponding to the change of the internal temperature and the pressure in the reactor container. Further, since all of these devices are disposed out of container, maintenance is also facilitated. (I.S.)

Reactor containing facility

International Nuclear Information System (INIS)

Akagawa, Katsuhiko.

1992-01-01

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

Reactor container

International Nuclear Information System (INIS)

Ichiki, Tadaharu; Saba, Kazuhisa.

1979-01-01

Purpose: To improve the earthquake resistance as well as reduce the size of a container for a nuclear reactor with no adverse effects on the decrease of impact shock to the container and shortening of construction step. Constitution: Reinforcing profile steel materials are welded longitudinally and transversely to the inner surface of a container, and inner steel plates are secured to the above profile steel materials while keeping a gap between the materials and the container. Reactor shielding wall planted to the base concrete of the container is mounted to the pressure vessel, and main steam pipeways secured by the transverse beams and led to the outside of container is connected. This can improve the rigidity earthquake strength and the safetiness against the increase in the inside pressure upon failures of the container. (Yoshino, Y.)

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)

Structural capacity assessment of WWER-1000 MW reactor containment. Progress report

International Nuclear Information System (INIS)

Jordanov, M.

1999-01-01

The objective of the project is to provide assessment of the structural behaviour and safety capacity of the WWER-1000 MW Reactor Building Containment at Kozloduy NPP under critical combination of loads according to the current international requirements. The analysis is focused on a realistic assessment of the Containment taking into account the non-linear shell behaviour of the pre-stressed reinforced concrete structure. Previous assessments of the status of pre stressing cables pointed out that the efficiency of the Containment as a final defence barrier for internal and external events depends on their reliability. Due to this, the experimental data obtained from embedded sensors (gauges) at pre-stressed shell structure is to be compared with the results from analytical investigations. The reliability of the WWER-1000 MW accident prevention system is under evaluation in the project. The Soviet standard design WWER-1000 MW type units installed in Kozloduy NPP were originally designed for a Safe Shutdown Earthquake (SSE) with a peak ground acceleration (PGA) of 0.1g. The new site seismicity studies revealed that the seismic hazard for the site significantly exceeds the originally estimated and a Review Level Earthquake (RLE) anchored to PGA=0.20g was proposed for re-assessment of the structures and equipment at Kozloduy NPP. The scope of the study is a re-assessment of the Containment structure under critical combination of loads according to the current safety and reliability requirements, including comparison between the Russian design requirements and the international regulations. Additionally, an investigation of the pre-stressing technology and the annual control of the cables' pre-stressing of the Containment is to be made. The crane influence on the dynamic behaviour of the Containment will be done as well as a study of the integrity of the Containment as a final defence barrier

Nuclear reactor containing facility

International Nuclear Information System (INIS)

Hidaka, Masataka; Murase, Michio.

1994-01-01

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

A method of installing a reactor container

International Nuclear Information System (INIS)

Hayashi, Kenji; Murakawa, Hisao.

1975-01-01

Object: To achieve exact installation of a reactor container at a site. Structure: A pole is set upright at the center of a cylindrical base portion, a plurality of beams are disposed around the pole in a radial fashion to form a cone, a plurality of steel plates are mounted successively around the cone through a ring, and the steel plates are welded to each other to assemble and install a reactor container at the same time. (Kamimura, M.)

Nuclear reactor container

International Nuclear Information System (INIS)

Moriyama, Takeo; Ochiai, Kanehiro; Niino, Tsuyoshi; Kodama, Toyokazu; Hirako, Shizuka.

1988-01-01

Purpose: To obtain structures suitable to a container structures for nuclear power plants used in those districts where earthquakes occur frequently, in which no local stresses are caused to the fundamental base portions and the workability for the fundamental structures is improved. Constitution: Basic stabilizers are attached to a nuclear reactor container (PCV) and a basic concrete recess for receiving a basic stabilizer is disposed in basic concretes. A top stabilizer is joined and fixed to a top stabilizer receiving plate at the inside of an outer shielding wall. On the other hand, a PCV top recess for conducting the load of PCV to the top stabilizer is attached to the top of the PCV. By disposing stabilizer structures allowing miner displacements at the two points, that is, the top and the lowermost portion of the PCV, no local stress concentrations can be generated to the extension on the axial direction of components due to the inner pressure of the PCV and to the horizontal load applied to the upper portion of the PCV upon earthquakes. (Yoshino, Y.)

The inner containment of an EPR trademark pressurized water reactor

Energy Technology Data Exchange (ETDEWEB)

Ostermann, Dirk; Krumb, Christian; Wienand, Burkhard [AREVA GmbH, Offenbach (Germany)

2014-08-15

On February 12, 2014 the containment pressure and subsequent leak tightness tests on the containment of the Finnish Olkiluoto 3 EPR trademark reactor building were completed successfully. The containment of an EPR trademark pressurized water reactor consists of an outer containment to protect the reactor building against external hazards (such as airplane crash) and of an inner containment that is subjected to internal overpressure and high temperature in case of internal accidents. The current paper gives an overview of the containment structure, the design criteria, the validation by analyses and experiments and the containment pressure test.

Some aspects of the reliability-based design of reactor containment structures

International Nuclear Information System (INIS)

Schueller, G.I.

1975-01-01

It is generally recognized that the load which a structure is likely to experience during its design life as well as its resistance are to be represented by random variables. A rational design procedure for reactor containment structures can therefore only be carried out within a probabilistic framework. Internal load conditions caused by system failure such as loss-of-coolant accident, pressure loads etc., and external load conditions caused for instance by impact due to aircraft crashes, external pressure waves and natural hazards such as earthquakes, floods, hurricanes are described by extreme value distributions of the Fisher-Tippett types. Statistical and physical arguments are given to support their application. The occurrence of these rare events with respect to time is modeled by a Poisson process. The yield strength of the containment structure for both steel (liner) and reinforced concrete shells is also modeled by extreme value distributions (of the smallest values). The failure criterion considered here is that of collapse determined by plastic yieldline formation. A failure mechanism as considered here describes a particular regime of plastic line formation. The probability of failure of a structure under a single load application of load types likely to occur during the design life of the structure is to be determined by integrating over all possible mechanisms. Finally Freudenthal's reliability function is utilized to combine the information derived above so that a containment design for given design lifes and reliabilities is possible. (orig.) [de

Structural response of rectilinear containment to overpressurization

International Nuclear Information System (INIS)

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

1995-01-01

Containment structures for nuclear reactors are the final barrier between released radionuclides and the public. Containment structures are constructed from steel, reinforced concrete, or prestressed concrete. US nuclear reactor containment geometries tend to be cylindrical with elliptical or hemispherical heads. The older Soviet designed reactors do not use a containment building to mitigate the effects of accidents. Instead, they employ a sealed set of rectilinear, interconnected compartments, collectively called the accident localization system (ALS), to reduce the release of radionuclides to the atmosphere during accidents. The purpose of this paper is to present a methodology that can be used to find the structural capacity of reinforced concrete structures. The method is applicable to both cylindrical and rectilinear geometries. As an illustrative example, the methodology is applied to a generic VVER-440/V213 design

Nuclear reactor container

International Nuclear Information System (INIS)

Yamaki, Rika; Kawabe, Ryuhei.

1989-01-01

A venturi scrubber is connected to a nuclear reactor container. Gases containing radioactive aerosols in the container are introduced into the venturi scrubber in the form of a high speed stream under the pressure of the container. The radioactive aerosols are captured by inertia collision due to the velocity difference between the high speed gas stream and water droplets. In the case of the present invention, since the high pressure of the reactor container generated upon accident is utilized, compressor, etc. is no more required, thereby enabling to reduce the size of the aerosol removing device. Further, since no external power is used, the radioactive aerosols can be removed with no starting failure upon accidents. (T.M.)

Emergency reactor container cooling facility

International Nuclear Information System (INIS)

Suzuki, Hiroaki; Matsumoto, Tomoyuki.

1992-01-01

The present invention concerns an emergency cooling facility for a nuclear reactor container having a pressure suppression chamber, in which water in the suppression chamber is effectively used for cooling the reactor container. That is, the lower portion of a water pool in the pressure suppression chamber and the inside of the reactor container are connected by a pipeline. The lower end of the pipeline and a pressurized incombustible gas tank disposed to the outside of the reactor container are connected by a pipeline by way of valves. Then, when the temperature of the lower end of the pressure vessel exceeds a predetermined value, the valves are opened. If the valves are opened, the incombustible gas flows into the lower end of the pipeline connecting the lower portion of the water pool in the pressure suppression chamber and the inside of the reactor container. Since the inside of the pipeline is a two phase flow comprising a mixture of a gas phase and a liquid phase, the average density is decreased. Therefore, the water level of the two phase flow is risen by the level difference between the inside and the outside of the pipeline and, finally, the two phase mixture is released into the reactor container. As a result, the reactor container can be cooled by water in the suppression chamber by a static means without requiring pumps. (I.S.)

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.)

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

Structural and containment response to LMFBR accidents

International Nuclear Information System (INIS)

Marchaterre, J.F.; Fistedis, S.H.; Baker, L. Jr.; Stepnewski, D.D.; Peak, R.D.; Gluekler, E.L.

1978-01-01

The results of current developments in analysing the response of reactor structures and containment to LMFBR accidents are presented. The current status of analysis of the structural response of LMFBR's to core disruptive accidents, including head response, potential missile generation and the effects of internal structures are presented. The results of recent experiments to help clarify the thermal response of reactor structures to molten core debris are summarized, including the use of this data to calculate the response of the secondary containment. (author)

Containment design, performance criteria and research needs for advanced reactor designs

International Nuclear Information System (INIS)

Bagdi, G.; Ali, S.; Costello, J

2004-01-01

This paper points out some important shifts in the basic expectations in the performance requirements for containment structures and discusses the areas where the containment structure design requirements and acceptance criteria can be integrated with ultimate test based insights. Although there has not been any new reactor construction in the United States for over thirty years, several designs of evolutionary and advanced reactors have already been certified. Performance requirements for containment structures under design basis and severe accident conditions and explicit consideration of seismic margins have been used in the design certification process. In the United States, the containment structure design code is the American Society of Mechanical Engineers, Boiler and Pressure Vessel Code, Section III, Division 1, Subsection NE-Class MC for the steel containment and Section III, Division 2 for reinforced and prestressed concrete reactor vessels and containments. This containment design code was based on the early concept of applying design basis internal pressure and associated load combinations that included the operating basis and safe shutdown earthquake ground motion. These early design criteria served the nuclear industry and the regulatory authorities in maintaining public health and safety. However, these early design criteria do not incorporate the performance criteria related to containment function in an integrated fashion. Research in large scale model testing of containment structures to failure from over pressurization and shake table testing using simulated ground motion, have produced insights related to failure modes and material behavior at failure. The results of this research provide the opportunity to integrate these observations into design and acceptance criteria. This integration process would identify 'gaps' in the present knowledge and future research needs. This knowledge base is important for gleaning risk-informed insights into

Passive pH adjustment of nuclear reactor containment flood water

International Nuclear Information System (INIS)

Gerlowski, T.J.

1986-01-01

A method is described of automatically and passively adjusting the pH of the recirculating liquid used to flood the containment structure of a nuclear reactor upon the occurence of an accident in order to cool the reactor core, wherein the containment structure has a concrete floor which is provided with at least one sump from which the liquid is withdrawn for recirculation via at least one outlet pipe. The method consists of: prior to flooding and during or prior to normal operation of the reactor, providing at least one perforated basket within at least one sump with the basket containing crystals of a pH adjusting chemical which is soluble in the liquid, and covering each basket with a plastic coating which is likewise soluble in the liquid, whereby upon flooding of the containment structure the liquid in the sump will reach the level of the baskets, causing the coating and the crystals to be dissolved and the chemical to mix with the recirculating liquid to adjust the pH

Reactor containment and reactor safety in the United States

International Nuclear Information System (INIS)

Kouts, H.

1986-01-01

The reactor safety systems of two reactors are studied aiming at the reactor containment integrity. The first is a BWR type reactor and is called Peachbottom 2, and the second is a PWR type reactor, and is called surry. (E.G.) [pt

Reactor container facility

International Nuclear Information System (INIS)

Saito, Takashi; Nagasaka, Hideo.

1990-01-01

A dry-well pool for spontaneously circulating stored pool water and a suppression pool for flooding a pressure vessel by feeding water, when required, to a flooding gap by means of spontaneous falling upto the flooding position, thereby flooding the pressure vessel are contained at the inside of a reactor container. Thus, when loss of coolant accidents such as caused by main pipe rupture accidents should happen, pool water in the suppression pool is supplied to the flooding gap by spontaneously falling. Further, if the flooding water uprises exceeding a predetermined level, the flooding gap is in communication with the dry-well pool at the upper and the lower portions respectively. Accordingly, flooding water at high temperature heated by the after-heat of the reactor core is returned again into the flooding gap to cool the reactor core repeatedly. (T.M.)

Mobile nuclear reactor containment vessel

International Nuclear Information System (INIS)

Thompson, R.E.; Spurrier, F.R.; Jones, A.R.

1978-01-01

A containment vessel for use in mobile nuclear reactor installations is described. The containment vessel completely surrounds the entire primary system, and is located as close to the reactor primary system components as is possible in order to minimize weight. In addition to being designed to withstand a specified internal pressure, the containment vessel is also designed to maintain integrity as a containment vessel in case of a possible collision accident

Evolution of nuclear reactor containments in India: Addressing the present day challenges

Energy Technology Data Exchange (ETDEWEB)

Kakodkar, Anil, E-mail: kakodkar@barc.gov.in

2014-04-01

Indigenously developed Pressurized Heavy Water Reactors (PHWRs) that form the backbone of current stage of nuclear power development in India have seen continuous evolution of their containment systems. This evolution that has taken place over implementation of 18 PHWRs (200/220/540 MWe) has encompassed all aspects of containment design, viz. the structural system, energy management system, radio-activity management and hydrogen management system. As a part of ongoing efforts toward strengthening of safety performance, India is also ready with the design of Advance Heavy Water Reactor (AHWR), which represents a technology demonstrator for advanced reactor systems and for thorium utilization. This reactor has a number of improved passive safety features and it is capable of meeting the demanding safety challenges that future reactor system would be expected to meet as a result of emerging expectations in the background of accidents over the past three decades viz. those at Three Mile Island (1979), Chernobyl (1986) and most recently at Fukushima (2011). In this lecture I shall focus on the evolution of nuclear reactor containments in India and highlight the design, associated structural and thermal hydraulics safety assessment made over the years for the improvement of containment performance.

Description of the containment for a stationary pressurized water reactor

International Nuclear Information System (INIS)

Hermani, S.

1986-01-01

The function of the containment is to prevent the inadvertent release of radioactive fission products from the reactor coolant system to the atmosphere and to provide biological shielding during both normal and accident operation. Basically three different containment concepts 1) the dry containment, 2) the subatmospheric containment, and 3) the ice condenser containment, have been developed, based on how the accident energy release from the reactor coolant system is controlled. The containment structure can be either 1) reinforced concrets with inside liner, 2) prestressed concrete with inside, or 3) full steel cylinder or steel sphere with separate concrete shield. The size of the containment is largely dictated by the required net free volume, that satisfies the energy release criteria due to the design basic accident. The design and construction methods applied to this structure guarantee that the containment will carry out its safety function. This was proven by the Three Mile Island accident. (author)

Liquid metal reactor applications of the CONTAIN code

International Nuclear Information System (INIS)

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

1988-01-01

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

Reactor container

International Nuclear Information System (INIS)

Hidaka, Masataka; Hatamiya, Shigeo; Kawasaki, Terufumi; Fukui, Toru; Suzuki, Hiroaki; Kataoka, Yoshiyuki; Kawabe, Ryuhei; Murase, Michio; Naito, Masanori.

1990-01-01

In order to suppress the pressure elevation in a reactor container due to high temperature and high pressure steams jetted out upon pipeway rupture accidents in the reactor container, the steams are introduced to a pressure suppression chamber for condensating them in stored coolants. However, the ability for suppressing the pressure elevation and steam coagulation are deteriorated due to the presence of inactive incondensible gases. Then, there are disposed a vent channel for introducing the steams in a dry well to a pressure suppression chamber in the reactor pressure vessel, a closed space disposed at the position lower than a usual liquid level, a first channel having an inlet in the pressure suppression chamber and an exit in the closed space and a second means connected by way of a backflow checking means for preventing the flow directing to the closed space. The first paths are present by plurality, a portion of which constitutes a syphon. The incondensible gases and the steams are discharged to the dry well at high pressure by using the difference of the water head for a long cooling time after the pipeway rupture accident. Then, safety can be improved without using dynamic equipments as driving source. (N.H.)

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.)

Vibration-damping structure for reactor building

International Nuclear Information System (INIS)

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

1998-01-01

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

Recent results on PEC reactor HCDA containment investigations

International Nuclear Information System (INIS)

Cenerini, R.; Palamidessi, A.; Verzelletti, G.

1979-01-01

The response of PEC reactor containement structures and of tank supporting arms to HCDA has been investigated by an explosive test on a refined 1:6 scaled mock-up. Experimental strains and pressures are compared with Astarte code calculations. (orig.)

Reinforced concrete containment structures in high seismic zones

International Nuclear Information System (INIS)

Aziz, T.S.

1977-01-01

A new structural concept for reinforced concrete containment structures at sites where earthquake ground motions in terms of the Safe Shutdown Earthquake (SSE) exceeds 0.3 g is presented. The structural concept is based on: (1) an inner steel-lined concrete shell which houses the reactor and provides shielding and containment in the event of loss of coolant accident; (2) an outer annular concrete shell structure which houses auxiliary reactor equipment and safeguards systems. These shell structures are supported on a common foundation mat which is embedded in the subgrade. Under stipulated earthquake conditions the two shell structures interact to resist lateral inertia forces. Thus the annular structure which is not a pressure boundary acts as a lateral support for the inner containment shell. The concept is practical, economically feasible and new to practice. (Auth.)

Cooling device for reactor container

International Nuclear Information System (INIS)

Akiba, Miyuki.

1996-01-01

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

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

International Nuclear Information System (INIS)

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

1997-12-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

1997-12-01

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

Conceptual design and technology development of containment structure in Fusion Experimental Reactor (FER)

International Nuclear Information System (INIS)

Nishio, Satoshi; Sato, Keisuke; Matsuoka, Fushiki; Kanamori, Naokazu; Koizumi, Koichi; Abe, Tetsuya; Hosobuchi, Hideo; Tada, Eisuke; Yamada, Masao.

1991-05-01

A conceptual design of FER (Fusion Experimental Reactor) containment structure and its associated R and D activities, conducted from '89 to '90, are described. The FER containment structure system which mainly consists of a vacuum vessel, shielding structures, in-vessel replaceable components, ports, a cooling pipe system, has been developed to fullfil the required function. As an initial stage of R and D activities, the elemental technologies common to a tokamak reactor have been developed. Among them, a locking mechanism for supporting in-vessel replaceable components and a technique for insulation/conduction are described. For the locking mechanism, a caulking cotter driven by hydraulic pressure has been employed. Three kinds of hydraulic driving mechanism have been manufactured by trial: a 'piston jack' type, a 'bellows' type and a 'flexible tube' type. In the latter type, the stroke is obtained by changing the cross section of the flexible tube from a flat racetrack shape to a fat shape by hydraulic pressure. As the result of preliminary performance test, the shape of 'flexible tube' has been found to be improved. For the insulation coating, Al 2 O 3 has been selected as the material and a plasma spray method has been applied as the coating procedure. For the conduction coating, Cr 3 C 2 has been selected as the material and JET-KOTE method has been applied as the coating procedure. Both methods have been successfully developed and have been confirmed to be applicable the actual machine. A one fifth scale model has been fabricated in order to verify the design feasibility, mainly geometrical consistency. Then some design modifications were found to be needed for some of the components based on the manufacturing experience. (author)

Nuclear reactor container

International Nuclear Information System (INIS)

Hosaka, Seiichi.

1988-01-01

Cables coverd with non-halogen covering material are used as electric wire cables wired for supplying electric power to a reactor recycling pump. Silicone rubber having specified molecular formula is used for the non-halogen covering material. As a result, formation of chlorine in a nuclear reactor container can be eliminated and increase in the deposited salts to SUS pipeways, etc. can be prevented, to avoid the occurrence of stress corrosion cracks. (H.T.)

Design of reactor containment systems for nuclear power plants. Safety guide

International Nuclear Information System (INIS)

2008-01-01

This Safety Guide was prepared under the IAEA programme for safety standards for nuclear power plants. It is a revision of the Safety Guide on Design of the Reactor Containment Systems in Nuclear Power Plants (Safety Series No. 50-Sg-D1) issued in 1985 and supplements the Safety Requirements publication on Safety of Nuclear Power Plants: Design. The present Safety Guide was prepared on the basis of a systematic review of the relevant publications, including the Safety of Nuclear Power Plants: Design, the Safety fundamentals publication on The Safety of Nuclear Installations, Safety Guides, INSAG Reports, a Technical Report and other publications covering the safety of nuclear power plants. 1.2. The confinement of radioactive material in a nuclear plant, including the control of discharges and the minimization of releases, is a fundamental safety function to be ensured in normal operational modes, for anticipated operational occurrences, in design basis accidents and, to the extent practicable, in selected beyond design basis accidents. In accordance with the concept of defence in depth, this fundamental safety function is achieved by means of several barriers and levels of defence. In most designs, the third and fourth levels of defence are achieved mainly by means of a strong structure enveloping the nuclear reactor. This structure is called the 'containment structure' or simply the 'containment'. This definition also applies to double wall containments. 1.3. The containment structure also protects the reactor against external events and provides radiation shielding in operational states and accident conditions. The containment structure and its associated systems with the functions of isolation, energy management, and control of radionuclides and combustible gases are referred to as the containment systems

Design of reactor containment systems for nuclear power plants. Safety guide

International Nuclear Information System (INIS)

2004-01-01

This Safety Guide was prepared under the IAEA programme for safety standards for nuclear power plants. It is a revision of the Safety Guide on Design of the Reactor Containment Systems in Nuclear Power Plants (Safety Series No. 50-Sg-D1) issued in 1985 and supplements the Safety Requirements publication on Safety of Nuclear Power Plants: Design. The present Safety Guide was prepared on the basis of a systematic review of the relevant publications, including the Safety of Nuclear Power Plants: Design, the Safety fundamentals publication on The Safety of Nuclear Installations, Safety Guides, INSAG Reports, a Technical Report and other publications covering the safety of nuclear power plants. 1.2. The confinement of radioactive material in a nuclear plant, including the control of discharges and the minimization of releases, is a fundamental safety function to be ensured in normal operational modes, for anticipated operational occurrences, in design basis accidents and, to the extent practicable, in selected beyond design basis accidents. In accordance with the concept of defence in depth, this fundamental safety function is achieved by means of several barriers and levels of defence. In most designs, the third and fourth levels of defence are achieved mainly by means of a strong structure enveloping the nuclear reactor. This structure is called the 'containment structure' or simply the 'containment'. This definition also applies to double wall containments. 1.3. The containment structure also protects the reactor against external events and provides radiation shielding in operational states and accident conditions. The containment structure and its associated systems with the functions of isolation, energy management, and control of radionuclides and combustible gases are referred to as the containment systems

Water injection device for reactor container

International Nuclear Information System (INIS)

Sakaki, Isao.

1996-01-01

A pressure vessel incorporating a reactor core is placed and secured on a pedestal in a dry well of a reactor container. A pedestal water injection line is disposed opened at one end in a pedestal cavity passing through the side wall of the pedestal and led at the other end to the outside of the reactor container. A substitution dry well spray line is connected to a spray header disposed at the upper portion of the dry well. When the pressure vessel should be damaged by a molten reactor core and the molten reactor core should drop to the dry well upon occurrence of an accident, the molten reactor core on the floor of the pedestal is cooled by water injection from the pedestal water injection line. At the same time, the elevation of the pressure and the temperature in the reactor container is suppressed by the water injection of the substitution dry well spray line. This can avoid large scaled release of radioactive materials to the environmental circumference. (I.N.)

Leak-tightness characteristics concerning the containment structures of the HTTR

International Nuclear Information System (INIS)

Sakaba, Nariaki; Iigaki, Kazuhiko; Kondo, Masaaki; Emori, Koichi

2004-01-01

The containment structures of the HTTR consist of the reactor containment vessel, the service area, and the emergency air purification system, which minimise the release of fission products in postulated accidents, which lead to fission product release from the reactor facilities. The reactor containment vessel is designed to withstand the temperature and pressure transients and to be leak-tight in the case of a rupture of the primary concentric hot-gas duct, etc. The pressure inside the service area is maintained at a negative pressure by the emergency air purification system. The emergency air purification system will also remove airborne radioactivity and will maintain a correct pressure in the service area. The leak-tightness characteristics of the containment structures are described in this paper. The measured leakage rates of the reactor containment vessel were enough less than the specified leakage limit of 0.1%/d confirmed during the commissioning tests and annual inspections. The service area was kept in a way that the design pressure becomes well below its allowable limitation by the emergency air purification system, which filters efficiency of particle removal and iodine removal well over the limited values. The obtained data demonstrate that the reactor containment structures were fabricated to minimise the release of fission products in the postulated accidents with fission product release from the reactor facilities

The WINCON programme - validation of fast reactor primary containment codes

International Nuclear Information System (INIS)

Sidoli, J.E.A.; Kendall, K.C.

1988-01-01

In the United Kingdom safety studies for the Commercial Demonstration Fast Reactor (CDFR) include an assessment of the capability of the primary containment in providing an adequate containment for defence against the hazards resulting from a hypothetical Whole Core Accident (WCA). The assessment is based on calculational estimates using computer codes supported by measured evidence from small-scale experiments. The hydrodynamic containment code SEURBNUK-EURDYN is capable of representing a prescribed energy release, the sodium coolant and cover gas, and the main containment and safety related internal structures. Containment loadings estimated using SEURBNUK-EURDYN are used in the structural dynamic code EURDYN-03 for the prediction of the containment response. The experiments serve two purposes, they demonstrate the response of the CDFR containment to accident loadings and provide data for the validation of the codes. This paper summarises the recently completed WINfrith CONtainment (WINCON) experiments that studied the response of specific features of current CDFR design options to WCA loadings. The codes have been applied to some of the experiments and a satisfactory prediction of the global response of the model containment is obtained. This provides confidence in the use of the codes in reactor assessments. (author)

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

Non-linear dynamic response of reactor containment

International Nuclear Information System (INIS)

Takemori, T.; Sotomura, K.; Yamada, M.

1975-01-01

A computer program was developed to investigate the elasto-plastic behavior of structures. This program is outlined and the problems of non-linear response of structures are discussed. Since the mode superposition method is only valid in an elastic analysis, the direct integration method was adopted here. As the sample model, an actual reactor containment (reactor building) of PWR plant was adopted. This building consists of three components, that is, a concrete internal structure, a steel containment vessel and a concrete outer shield wall. These components are resting on a rigid foundation mat. Therefore they were modeled with a lumped mass model respectively and coupled on the foundation. The following assumptions were employed to establish the properties of dynamic model: rocking and swaying springs of soil can be obtained from an elastic half-space solution, and the hysteretic characteristic of springs is bi-linear; springs connecting each mass are dealt with shear beams so that both bending and shear deflections can be included (Hysteretic characteristics of springs are linear, bi-linear and tri-linear for the internal structure, the containment vessel and the outer shield wall, respectively); generally, each damping coefficient is given for each mode in modal superposition (However, a damping matrix must be made directly in a non-linear response). Therefore the damping matrix of the model was made by combining the damping matrices [C] of each component obtained by Caughy's method and a damping value of the rocking and swaying by the half-space solution. On the basis of above conditions, the non-linear response of the structure was obtained and the difference between elastic and elasto-plastic analysis is presented

Bellefonte primary containment structure

International Nuclear Information System (INIS)

Olyniec, J.H.

1981-01-01

Construction of the reactor building primary containment structure at the Bellefonte Nuclear Plant involved several specialized construction techniques. This two unit plant is one of the nine nuclear units at six different sites now under construction by the Tennessee Valley Authority (TVA). The post-Tensioned, cast-in-place interior steel lined containment structure is unique within TVA. Problems during construction were identified at weekly planning meetings, and options were discussed. Close coordination between craft supervisors and on-site engineering personnel drew together ''hands-on''experience and technical background. Details of the construction techniques, problems, and solutions are presented

Cooling device for reactor container

International Nuclear Information System (INIS)

Arai, Kenji.

1996-01-01

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

Dynamic analysis of reactor containment building using axisymmetric finite element model

International Nuclear Information System (INIS)

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

1989-01-01

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

Nuclear reactor fuel structure containing uranium alloy wires embedded in a metallic matrix plate

International Nuclear Information System (INIS)

Travelli, A.

1988-01-01

A nuclear fuel-containing plate structure for a nuclear reactor is described; such structure comprising a pair of malleable metallic non-fissionable matrix plates having confronting surfaces which are pressure bonded together and fully united to form a bonded surface, and elongated malleable wire-like fissionable fuel members separately confined and fully enclosed between the matrix plates along the interface to afford a high fuel density as well as structural integrity and effective retention of fission products. The plates have separate recesses formed in the confronting surfaces for closely receiving the wire-like fissionable fuel members. The wire-like fissionable fuel members are made of a maleable uranium alloy capable of being formed into elongated wire-like members and capable of withstanding pressure bonding. The wire-like fissionable fuel members are completely separated and isolated by fully united portions of the interface

Development of a fresh plutonium fuel container for a prototype fast breeder reactor

International Nuclear Information System (INIS)

Ohtake, T.; Takahashi, S.; Mishima, T.; Kurakami, J.; Yamamoto, Y.; Ohuchi, Y.

1989-01-01

Japan gives a good deal of encouragement to development of a fast breeder reactor (which is considered as the most likely candidate for nuclear power generation) to secure long-term energy source. And, following an experimental fast breeder reactor Joyo, a prototype fast breeder reactor Monju is now under vigorous construction. Related to development of the prototype fast breeder reactor, it is necessary and important to develop transport container which is used for transporting fresh fuel assemblies from Plutonium Fuel Production Facility to the Monju power plant. Therefore, the container is now being developed by Power Reactor and Nuclear Fuel Development Corporation (PNC). Currently, shipment and vibration tests, handling performance tests, shielding performance tests and prototype container tests are executed with prototype containers fabricated according to a final design, in order to experimentally confirm soundness of transport container and its contents, and propriety of design technique. This paper describes the summary of general specifications and structures of this container and the summary of preliminary safety analysis of package

Reactor container and controlling method thereof

International Nuclear Information System (INIS)

Hosaka, Seiichi.

1990-01-01

An object of the present invention is to prevent stress corrosion crack caused in pipelines made of stainless steels by preventing deposition of chlorine, etc. on the surface, etc. of the pipelines. That is, an internal evolving gas elimination system comprises a gas extraction device for extracting gases in the reactor container, an obstacle elimination device for eliminating obstacles contained in the extracted gases and an internal gas elimination device for eliminating internal evolving gases contained in the extracted gases. Further, gases in the upper portion of the reactor container are extracted and then the ingredients of the internal evolving gases contained in the gases are eliminated and, thereafter, the gases are supplied to the lower portion of the container to keep the relative humidity in the reactor container to less than 20%RH. As a result, since the internal evolving gases are eliminated and the relative humidity at the inside is kept to less than 20%RH, deposition of chlorine or salts on the pipelines can be prevented to thereby prevent the stress corrosion cracks. (I.S.)

Containment for small pressurized water reactors

International Nuclear Information System (INIS)

Siler, W.C.; Marda, R.S.; Smith, W.R.

1977-01-01

Babcock and Wilcox Company has prepared studies under ERDA contract of small and intermediate size (313, 365 and 1200 MWt) PWR reactor plants, for industrial cogeneration or electric power generation. Studies and experience with nuclear plants in this size range indicate unfavorable economics. To offset this disadvantage, modular characteristics of an integral reactor and close-coupled vapor suppression containment have been exploited to shorten construction schedules and reduce construction costs. The resulting compact reactor/containment complex is illustrated. Economic studies to date indicate that the containment design and the innovative construction techniques developed to shorten erection schedules have been important factors in reducing estimated project costs, thus potentially making such smaller plants competetive with competing energy sources

Assessment of factors that may affect the moisture- and temperature variations in the concrete structures inside nuclear reactor containments

International Nuclear Information System (INIS)

Oxfall, M.; Hassanzadeh, M.; Johansson, P.

2015-01-01

Three factors that may affect the climatic conditions inside Swedish nuclear reactor containments are the outdoor climate, the cooling water temperature and the operational state of the reactor. Those factors that have a considerable affect on the climatic conditions will be identified through comparisons to actual conditions during operation inside reactor containments. Knowledge about the impact of the factors on the climatic conditions is essential when developing a model to determine the prevailing and to predict future conditions in the concrete structures. The comparison between a Pressurized Water Reactor and a Boiling Water Reactor showed that there is no clear similarity, with regard to fluctuation of the indoor climate conditions, between the two reactor types. The indoor climate at a Pressurized Water Reactor seemed not to be affected of changes in the operational state, but it follows the fluctuations of the outdoor temperature and to some extent the water temperature. The Boiling water Reactor did not follow the water or outdoor temperature fluctuations. However, at a full power outage during a short period of time during the operational year the temperature drastically decreased. An operational year of a Boiling water reactor can be divided into three periods, where the first represents the yearly power outage, the second represents the autumn-winter-spring period and the third represents the summer period. The operational year of a pressurized water reactor can't easily be divided into stable periods, since the indoor temperature fluctuation follows the outdoor temperature fluctuation. Based on these measurements a simplified model which includes outer factors is possible for a BWR but more difficult for a PWR. (authors)

Light and heavy water replacing system in reactor container

International Nuclear Information System (INIS)

Miyamoto, Keiji.

1979-01-01

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

A model to predict moisture conditions in concrete reactor containments

International Nuclear Information System (INIS)

Ahs, M.; Nilsson, L.O.; Poyet, S.; L'Hostis, V.

2015-01-01

Moisture has an impact in many of the degradation mechanisms that appear in the structures of a nuclear power plant. Moisture conditions in a reactor containment wall have been simulated by using a hygro-thermal model of drying concrete. Methods to estimate the temperature dependency of the sorption isotherms and moisture transport properties is suggested and applied in the model. This temperature dependency is included as there is a temperature gradient present through the containment wall. The hygro-thermal model was applied on a full scale 3D model of a real reactor containment building and the concrete relative humidity has been computed at 4 different moments: 1, 10, 20 and 30 years. The results show that the major part of the concrete is not dried at all even after 30 years of operation. It is also clear that the temperature distribution inside the whole concrete volume is affected by the variable boundary conditions. It was concluded that the suggested hygro-thermal model was appropriate to use as a method to estimate the existing conditions in a PWR reactor containment wall

Leak monitoring method for a reactor container

International Nuclear Information System (INIS)

Uehara, Toshio.

1987-01-01

Purpose: To confirm leakages from a container upon nuclear reactor operation. Method: Leakages from a nuclear reactor container has been prevented by lowering the inner pressure of the container relative to the external pressure. In the conventional method of calculating the leakage by applying an inner pressure to the container and measuring the pressure change, etc. after the elapse of a pre-determined time, the measurement has to be conducted at periodical inspection when the nuclear reactor is shut-down. In view of the above, the leak test is conducted in the present invention by applying a slight inner pressure to the inside of the reactor container by supplying gases from a gas supply system and detecting the flow rate of the gases in the gas supply system while maintaining the slight inner pressure constant by controlling the supply and discharge of the gases. By applying such a inner pressure as causing no effect to the reactor operation, it is possible to monitor the leaks during operation and to detect the flow rate value surely and continuously if the leak is resulted. (Kamimura, M.)

Life management for a non replaceable structure: the reactor building

International Nuclear Information System (INIS)

Torres, V.; Francia, L.

1998-01-01

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

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)

Seismic analysis of liquid storage container in nuclear reactors

International Nuclear Information System (INIS)

Zhang Zhengming; He Shuyan; Xu Ming

2007-01-01

Seismic analysis of liquid storage containers is always difficult in the seismic design of nuclear reactor equipment. The main reason is that the liquid will generate significant seismic loads under earthquake. These dynamic liquid loads usually form the main source of the stresses in the container. For this kind of structure-fluid coupling problem, some simplified theoretical methods were usually used previously. But this cannot satisfy the requirements of engineering design. The Finite Element Method, which is now full developed and very useful for the structural analysis, is still not mature for the structure-fluid coupling problem. This paper introduces a method suitable for engineering mechanical analysis. Combining theoretical analysis of the dynamic liquid loads and finite element analysis of the structure together, this method can give practical solutions in the seismic design of liquid storage containers

Ventilation air conditioner for a reactor container

International Nuclear Information System (INIS)

Ikegame, Noboru; Nakagawa, Takeshi.

1980-01-01

Purpose: To suppress the variations in the internal pressure of a reactor container and smoothly ventilate the reactor container. Constitution: The air conditioner provides an air-flow-rate-control damper, a purge-air supply fan, and a filter device in the air-supply pipe of a reactor container. Furthermore, it provides a pressure difference detector at a part of the container. The air-flow-rate-control damper is connected electrically through a position-modulator-comparison amplifier to the pressure difference detector. When the filtration becomes insufficient by clogging of the filter device and the internal pressure increased abruptly in the container, the pressure-difference detector can detect it, and the damper is operated by a pressure regulator and the comparator so as to control the air flow to the container. Thus, the internal pressure variation is controlled so as to easily ventilate the container. (J.P.N.)

Reactor container

International Nuclear Information System (INIS)

Oyamada, Osamu; Furukawa, Hideyasu; Uozumi, Hiroto.

1979-01-01

Purpose: To lower the position of an intermediate slab within a reactor container and fitting a heat insulating material to the inner wall of said intermediate slab, whereby a space for a control rod exchanging device and thermal stresses of the inner peripheral wall are lowered. Constitution: In the pedestal at the lower part of a reactor pressure vessel there is formed an intermediate slab at a position lower than diaphragm floor slab of the outer periphery of the pedestal thereby to secure a space for providing automatic exchanging device of a control rod driving device. Futhermore, a heat insulating material is fitted to the inner peripheral wall at the upper side of the intermediate slab part, and the temperature gradient in the wall thickness direction at the time of a piping rupture trouble is made gentle, and thermal stresses at the inner peripheral wall are lowered. (Sekiya, K.)

Pressure releasing device for reactor container

International Nuclear Information System (INIS)

Takeda, Mika.

1994-01-01

In the present invention, dose rate to public caused by radioactive rare gases can be decreased. That is, a reactor container contains a reactor pressure vessel incorporating a reactor core. There are disposed a pressure releasing system for releasing the pressure in the reactor pressure vessel to the outside, and a burning device for burning gases released from the pressure releasing system. An exhaustion pipe is disposed to the pressure releasing system. A burning device is disposed to the exhaustion pipe. It is effective to dispose a ventilation port at a portion of the exhaustion pipe upstream of the burning device. In addition, the burning device may preferably be disposed in a multi-stage in the axial direction of the exhaustion pipe. With such procedures, hydrogen in gases discharged along with the release of the pressure in the container is burned. Buoyancy is caused to the exhaustion gases by heat energy upon burning. Since the exhaustion gases can reach a higher level by the buoyancy, the dose rate due to the rare gases can be reduced. (I.S.)

Humidity control device in a reactor container

International Nuclear Information System (INIS)

Aizawa, Motohiro; Igarashi, Hiroo; Osumi, Katsumi; Kimura, Takashi.

1986-01-01

Purpose: To provide a device capable of maintaining the inside of a container under high humidity circumstantial conditions causing less atmospheric corrosions, in order to prevent injuries due to atmospheric corrosions to smaller diameter stainless steel pipeways in the reactor container. Constitution: Stress corrosion cracks (SCC) to the smaller diameter stainless steel pipeways are caused dependent on the relative humidity and it is effective as the countermeasure against SCC to maintain the relative humidity at a low level less than 30 % or high level greater than 60 %. Based on the above findings, a humidity control device is disposed so as to maintain the relative humidity for the atmosphere within a reactor core on a higher humidity region. The device is adapted such that recycling gas in the dry-well coolant circuit is passed through an orifice to atomize the water introduced from feedwater pipe and introduce into a reactor core or such that the recycling gases in the dry-well cooling circuit are bubbled into water to remove chlorine gas in the reactor container gas thereby increasing the humidity in the reactor container. (Kamimura, M.)

Trends in the development of reactor containments

International Nuclear Information System (INIS)

Turricchia, A.

1977-01-01

The paper presents a review of the technical developments which the containment systems of BWRs, PWRs and HWRs have undergone in the USA, Europe and Canada. The great variety of existing containment systems have been classified for each type of reactor, according to: principle of operation (i.e. dry containment; pressure suppression containment; containment with vacuum building; sub-atmospheric containment; etc.); type and number of the physical barriers provided against the dispersion of fission products to the environment (single containment, double containment, single containment with local collection of the leaks from the most probable leakage paths, or with pressurization of the potential leakage paths, etc.); and structural characteristics of the containment barriers (reinforced concrete, prestressed concrete, metal containment, liners). In the framework of this classification the main features of the various containment systems are illustrated, with special emphasis on the most modern designs. The difference in trend among various countries and various designers are highlighted. A review is also made of the various containment auxiliary systems which would operate in accident conditions (isolation systems, heat removal system, atmosphere control system, etc.). Also for these systems the main design criteria are set forth and present trends are described. Finally, a brief survey is also made of the present trends with regards to the design of containment structures against external events be they of natural origin (earthquake, tornadoes, etc.) or connected with human activities (airplane crash, explosions, etc.) [fr

Sensitivity Study on Analysis of Reactor Containment Response to LOCA

International Nuclear Information System (INIS)

Chung, Ku Young; Sung, Key Yong

2010-01-01

As a reactor containment vessel is the final barrier to the release of radioactive material during design basis accidents (DBAs), its structural integrity must be maintained by withstanding the high pressure conditions resulting from DBAs. To verify the structural integrity of the containment, response analyses are performed to get the pressure transient inside the containment after DBAs, including loss of coolant accidents (LOCAs). The purpose of this study is to give regulative insights into the importance of input variables in the analysis of containment responses to a large break LOCA (LBLOCA). For the sensitivity study, a LBLOCA in Kori 3 and 4 nuclear power plant (NPP) is analyzed by CONTEMPT-LT computer code

Sensitivity Study on Analysis of Reactor Containment Response to LOCA

Energy Technology Data Exchange (ETDEWEB)

Chung, Ku Young; Sung, Key Yong [Korea Institute of Nuclear Safety, Daejeon (Korea, Republic of)

2010-10-15

As a reactor containment vessel is the final barrier to the release of radioactive material during design basis accidents (DBAs), its structural integrity must be maintained by withstanding the high pressure conditions resulting from DBAs. To verify the structural integrity of the containment, response analyses are performed to get the pressure transient inside the containment after DBAs, including loss of coolant accidents (LOCAs). The purpose of this study is to give regulative insights into the importance of input variables in the analysis of containment responses to a large break LOCA (LBLOCA). For the sensitivity study, a LBLOCA in Kori 3 and 4 nuclear power plant (NPP) is analyzed by CONTEMPT-LT computer code

Reactor container spray device

International Nuclear Information System (INIS)

Yanai, Ryoichi.

1980-01-01

Purpose: To enable decrease in the heat and the concentration of radioactive iodine released from the reactor vessel into the reactor container in the spray device of BWR type reactors. Constitution: A plurality of water receiving trays are disposed below the spray nozzle in the dry well and communicated to a pressure suppression chamber by way of drain pipeways passing through a diaphragm floor. When the recycling system is ruptured and coolants in the reactor vessel and radioactive iodine in the reactor core are released into the dry well, spray water is discharged from the spray nozzle to eliminate the heat and the radioactive iodine in the dry well. In this case, the receiving trays collect the portions of spray water whose absorption power for the heat and radioactive iodine is nearly saturated and falls them into the pool water of the pressure suppression chamber. Consequently, other portions of the spray water that still possess absorption power can be jetted with no hindrance, to increase the efficiency for the removal of the heat and iodine of the spray droplets. (Horiuchi, T.)

CONCRETE REACTOR CONTAINMENT

Energy Technology Data Exchange (ETDEWEB)

Lumb, Ralph F.; Hall, William F.; Fruchtbaum, Jacob

1963-06-15

The results of various leak-rate tests demonstrate the practicality of concrete as primary containment for the maximum credible accident for a research reactor employing plate-type fuel and having a power in excess of one megawatt. Leak-test time was shortened substantially by measuring the relaxation time for overpressure decay, which is a function of leak rate. (auth)

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

Reactor container

International Nuclear Information System (INIS)

Furukawa, Hideyasu; Oyamada, Osamu; Uozumi, Hiroto.

1976-01-01

Purpose: To provide a container for a reactor provided with a pressure suppressing chamber pool which can prevent bubble vibrating load, particularly negative pressure generated at the time of starting to release exhaust from a main steam escape-safety valve from being transmitted to a lower liner plate of the container. Constitution: This arrangement is characterized in that a safety valve exhaust pool for main steam escape, in which a pressure suppressing chamber pool is separated and intercepted from pool water in the pressure suppressing chamber pool, a safety valve exhaust pipe is open into said safety valve exhaust pool, and an isolator member, which isolates the bottom liner plate in the pressure suppressing chamber pool from the pool water, is disposed on the bottom of the safety valve exhaust pool. (Nakamura, S.)

The role of nuclear reactor containment in severe accidents

International Nuclear Information System (INIS)

1989-04-01

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

Method of judging leak sources in a reactor container

International Nuclear Information System (INIS)

Maeda, Katsuji.

1984-01-01

Purpose: To enable exact judgement for leak sources upon leak accident in a reactor container of BWR type power plants as to whether the sources are present in the steam system or coolant system. Method: If leak is resulted from the main steam system, the hydrogen density in the reactor container is about 170 times as high as the same amount of leak from the reactor water. Accordingly, it can be judged whether the leak source is present in the steam system or reactor water system based on the change in the indication of hydrogen densitometer within the reactor container, and the indication from the drain amount from the sump in the container or the indication of a drain flow meter in the container dehumidifier. Further, I-131, Na-24 and the like as the radioactive nucleides in sump water of the container are measured to determine the density ratio R = (I-131)/(Na-24), and it is judged that the leak is resulted in nuclear water if the density ratio R is equal to that of reactor water and that the leak is resulted from the main steam or like other steam system if the density ratio R is higher than by about 100 times than that of reactor water. (Horiuchi, T.)

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

Nuclear reactor structural material forming less radioactive corrosion product

International Nuclear Information System (INIS)

Nakazawa, Hiroshi.

1988-01-01

Purpose: To provide nuclear reactor structural materials forming less radioactive corrosion products. Constitution: Ni-based alloys such as inconel alloy 718, 600 or inconel alloy 750 and 690 having excellent corrosion resistance and mechanical property even in coolants at high temperature and high pressure have generally been used as nuclear reactor structural materials. However, even such materials yield corrosion products being attacked by coolants circulating in the nuclear reactor, which produce by neutron irradiation radioactive corrosion products, that are deposited in primary circuit pipeways to constitute exposure sources. The present invention dissolves dissolves this problems by providing less activating nuclear reactor structural materials. That is, taking notice on the fact that Ni-58 contained generally by 68 % in Ni changes into Co-58 under irradiation of neutron thereby causing activation, the surface of nuclear reactor structural materials is applied with Ni plating by using Ni with a reduced content of Ni-58 isotopes. Accordingly, increase in the radiation level of the nuclear reactor structural materials can be inhibited. (K.M.)

Containment concepts assessment for the SEAFP reactor

International Nuclear Information System (INIS)

Di Pace, L.; Natalizio, A.

2000-01-01

A simple methodology has been developed for making relative comparisons of potential containment designs for future fusion reactors. The assessment methodology requires only conceptual design information. The application of this methodology, at the early stages of a fusion reactor design, provides designers useful information regarding the suitability of various containment designs and design features. Because the radiation hazard from the operation of future fusion power reactors is expected to be low, the containment design, in addition to public safety, needs to take into account worker safety considerations, as well as factors important to the reliable and economical operation of the power plant. Several containment concepts have been assessed with a methodology that takes into account public safety, worker safety, operability and maintainability as well as cost. This paper describes this methodology and presents the results of the assessment. The paper concludes that, to obtain a containment design that is optimised with respect to safety, operational and cost factors, designers should focus on a containment that is conceptually simple-that is, one utilising a single, large containment building without relying on special features such as expansion volumes, pressure suppression pools or spray systems

Containment Loads Analysis for CANDU6 Reactor using CONTAIN 2.0

International Nuclear Information System (INIS)

Kim, Tae H.; Yang, Chae Y.

2013-01-01

The containment plays an important role to limit the release of radioactive materials to the environment during design basis accidents (DBAs). Therefore, the containment has to maintain its integrity under DBA conditions. Generally, a containment functional DBA evaluation includes calculations of the key containment loads, i. e., pressure and temperature effects associated with a postulated large rupture of the primary or secondary coolant system piping. In this paper, the behavior of containment pressure and temperature was evaluated for loss of coolant accidents (LOCAs) of the Wolsong unit 1 in order to assess the applicability of CONTAIN 2.0 code for the containment loads analysis of the CANDU6 reactor. The containment pressure and temperature of the Wolsong unit 1 were evaluated using the CONTAIN 2.0 code and the results were compared with the CONTEMPT4 code. The peak pressure and temperature calculated by CONTAIN 2.0 agreed well with those of CONTEMPT4 calculation. The overall result of this analysis shows that the CONTAIN 2.0 code can apply to the containment loads analysis for the CANDU6 reactor

Excess-pressure suppression device in a reactor container

International Nuclear Information System (INIS)

Nishio, Masahide

1985-01-01

Purpose: To reliably decrease the radioactivity of radioactive gases when they are released externally. Constitution: The exit of a gas exhaust pipe for discharging gases in a reactor container, on generation of an excess pressure in the reactor container upon loss of coolant accident, is adapted to be always fluided in the cooling tank. Then, the exhaust gases discharged in the cooling tank is realeased to the atmosphere. In this way, the excess pressure in the reactor container can be prevented previously and the radioactivity of the gases released externally is significantly reduced by the scrubbing effect. (Kamimura, M.)

Vent control device for nuclear reactor container

International Nuclear Information System (INIS)

Kubota, Ryuji.

1989-01-01

The present invention concerns automatic prevention of abnormal over-pressure and hydrogen gas flashing in a BWR type reactor container. That is, (1) if the pressure in the container is abnormally increased, the gas in the pressure suppression chamber is released to reduce the pressure thereby preventing over-pressure damage to the container. (2) Then, if exhaust gases are burnt to cause flashing explosion danger for the gases in the reactor container, the gas release is interrupted. The foregoing two functioins are automatically conducted in this device. Specifically, when the pressure in the reactor container reaches a predetermined allowable limit, a remote control operation valve is opened by automatic control means to release the gas in the vessel. Since the gas flow rate at the start of the release exceeds flame propagation velocity, there is no worry for flashing explosion. Further, if the pipeway flow velocity near the atmospheric release is reduced to less than the flame propagation velocity of the hydrogen gas, the opened valve is automatically closed. Accordingly, propagation of hydrogen gas flame into the container thus causing explosion can surely be prevented. (K.M.)

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.)

Simplified safety and containment systems for the iris reactor

International Nuclear Information System (INIS)

Conway, L.E.; Lombardi, C.; Ricotti, M.; Oriani, L.

2001-01-01

The IRIS (International Reactor Innovative and Secure) is a 100 - 300 MW modular type pressurized water reactor supported by the U.S. DOE NERI Program. IRIS features a long-life core to provide proliferation resistance and to reduce the volume of spent fuel, as well as reduce maintenance requirements. IRIS utilizes an integral reactor vessel that contains all major primary system components. This integral reactor vessel makes it possible to reduce containment size; making the IRIS more cost competitive. IRIS is being designed to enhance reactor safety, and therefore a key aspect of the IRIS program is the development of the safety and containment systems. These systems are being designed to maximize containment integrity, prevent core uncover following postulated accidents, minimize the probability and consequences of severe accidents, and provide a significant simplification over current safety system designs. The design of the IRIS containment and safety systems has been identified and preliminary analyses have been completed. The IRIS safety concept employs some unique features that minimize the consequences of postulated design basis events. This paper will provide a description of the containment design and safety systems, and will summarize the analysis results. (author)

Structure of diaphragm floor of reactor container, construction module and construction method thereof

International Nuclear Information System (INIS)

Hasegawa, Hiroshi; Oikawa, Tadaaki; Ushiroda, Koichi; Matsuura, Tadashi; Komaru, Toshimi; Nemoto, Yoichi; Makita, Tatsuo; Maezawa, Sumito.

1998-01-01

A diaphragm floor of a reactor container has a structure comprising iron beams buried in concretes and connection members connecting the iron beams and liners, in which the liners are supported by the iron beams, and the load of the iron reinforced concretes when formed on the liner is supported by the iron beams thereby enabling to construct a diaphragm floor with no or reduced amount of temporary support members. As a result, the construction operation can be promoted by reducing the amount of the temporary support members or making the removing operation of the temporary support members unnecessary. The concrete layer comprises at least two upper and lower layers of a firstly formed concrete layer and a subsequently formed concrete layer, and the iron beams have such a strength capable of enduring the load applied when the firstly formed concrete is placed. (N.H.)

Reinforced concrete containment structures in high seismic zones

International Nuclear Information System (INIS)

Aziz, T.S.

1977-01-01

A new structural concept for reinforced concrete containment structures at sites where earthquake ground motions in terms of the Safe Shutdown Earthquake (SSE) exceeds 0.3 g is presented. The structural concept is based on: (1) an inner steel-lined concrete shell which houses the reactor and provides shielding and containment in the event of loss of coolant accident; (2) an outer annular concrete shell structure which houses auxilary reactor equipment and safeguards systems. These shell structures are supported on a common foundation mat which is embeded in the subgrade. Under stipulated earthquake conditions the two shell structures interact to resist lateral inertia forces. Thus the annular structure which is not a pressure boundary acts as a lateral support for the inner containment shell. The concept is practical, economically feasible and new to practice. An integrated configuration which includes the interior shell, the annular structure and the subgrade is analyzed for several static and dynamic loading conditions. The analysis is done using a finite difference solution scheme for the static loading conditions. A semi-analytical three-dimensional finite element scheme combined with a Fast Fourier Transform (FFT) algorithm is used for the dynamic loading conditions. The effects of cracking of the containment structure due to pressurization in conjunction with earthquake loading are discussed. Analytical results for both the finite difference and the finite element schemes are presented and the sensitivity of the results to changes in the input parameters is studied. General recommendations are given for plant configurations where high seismic loading is a major design consideration

Underground reactor containments: An option for the future?

International Nuclear Information System (INIS)

Forsberg, C.W.; Kress, T.

1997-01-01

Changing world conditions and changing technologies suggest that serious consideration should be given to siting of nuclear power plants underground. Underground siting is not a new concept. Multiple research reactors, several weapons production reactors, and one power reactor have been built underground. What is new are the technologies and incentives that may now make underground siting a preferred option. The conditions and technologies, along with their implications, are discussed herein. Underground containments can be constructed in mined cavities or pits that are then backfilled with thick layers of rock and soil. Conventional above-ground containments resist assaults and accidents because of the strength of their construction materials and the effectiveness of their safety features that are engineered to reduce loads. However, underground containments can provide even more resistance to assaults and accidents because of the inertia of the mass of materials over the reactor. High-technology weapons or some internal accidents can cause existing strong-material containments to fail, but only very-high energy releases can move large inertial masses associated with underground containments. New methods of isolation may provide a higher confidence in isolation that is independent of operator action

Simulation of hydrogen distribution in an Indian Nuclear Reactor Containment

Energy Technology Data Exchange (ETDEWEB)

Prabhudharwadkar, Deoras M. [Department of Mechanical Engineering, Indian Institute of Technology, Mumbai (India); Iyer, Kannan N., E-mail: kiyer@iitb.ac.i [Department of Mechanical Engineering, Indian Institute of Technology, Mumbai (India); Mohan, Nalini; Bajaj, Satinder S. [Nuclear Power Corporation of India Ltd., Mumbai (India); Markandeya, Suhas G. [Bhabha Atomic Research Centre, Trombay, Mumbai (India)

2011-03-15

Research highlights: This work addresses hydrogen dispersion in commercial nuclear reactor containment. The numerical tool used for simulation is first benchmarked with experimental data. Parametric results are then carried out for different release configurations. Results lead to the conclusion that the dispersal is buoyancy dominated. Also, the hydrogen concentration is high enough to demand mitigation devices. - Abstract: The management of hydrogen in a Nuclear Reactor Containment after LOCA (Loss Of Coolant Accident) is of practical importance to preserve the structural integrity of the containment. This paper presents the results of systematic work carried out using the commercial Computational Fluid Dynamics (CFD) software FLUENT to assess the concentration distribution of hydrogen in a typical Indian Nuclear Reactor Containment. In order to obtain an accurate estimate of hydrogen concentration distribution, a suitable model for turbulence closure is required to be selected. Using guidelines from the previous studies reported in the literature and a comparative simulation study using simple benchmark problems, the most suitable turbulence model for hydrogen mixing prediction was identified. Subsequently, unstructured meshes were generated to represent the containment of a typical Indian Nuclear Reactor. Analyses were carried out to quantify the hydrogen distribution for three cases. These were (1) Uniform injection of hydrogen for a given period of time at room temperature, (2) Time varying injection as has been computed from an accident analysis code, (3) Time varying injection (as used in case (2)) at a high temperature. A parametric exercise was also carried out in case (1) where the effect of various inlet orientations and locations on hydrogen distribution was studied. The results indicate that the process of hydrogen dispersal is buoyancy dominated. Further for typical injection rates encountered following LOCA, the dispersal is quite poor and most

Simulation of hydrogen distribution in an Indian Nuclear Reactor Containment

International Nuclear Information System (INIS)

Prabhudharwadkar, Deoras M.; Iyer, Kannan N.; Mohan, Nalini; Bajaj, Satinder S.; Markandeya, Suhas G.

2011-01-01

Research highlights: → This work addresses hydrogen dispersion in commercial nuclear reactor containment. → The numerical tool used for simulation is first benchmarked with experimental data. → Parametric results are then carried out for different release configurations. → Results lead to the conclusion that the dispersal is buoyancy dominated. → Also, the hydrogen concentration is high enough to demand mitigation devices. - Abstract: The management of hydrogen in a Nuclear Reactor Containment after LOCA (Loss Of Coolant Accident) is of practical importance to preserve the structural integrity of the containment. This paper presents the results of systematic work carried out using the commercial Computational Fluid Dynamics (CFD) software FLUENT to assess the concentration distribution of hydrogen in a typical Indian Nuclear Reactor Containment. In order to obtain an accurate estimate of hydrogen concentration distribution, a suitable model for turbulence closure is required to be selected. Using guidelines from the previous studies reported in the literature and a comparative simulation study using simple benchmark problems, the most suitable turbulence model for hydrogen mixing prediction was identified. Subsequently, unstructured meshes were generated to represent the containment of a typical Indian Nuclear Reactor. Analyses were carried out to quantify the hydrogen distribution for three cases. These were (1) Uniform injection of hydrogen for a given period of time at room temperature, (2) Time varying injection as has been computed from an accident analysis code, (3) Time varying injection (as used in case (2)) at a high temperature. A parametric exercise was also carried out in case (1) where the effect of various inlet orientations and locations on hydrogen distribution was studied. The results indicate that the process of hydrogen dispersal is buoyancy dominated. Further for typical injection rates encountered following LOCA, the dispersal is

Heat removing device for reactor container

International Nuclear Information System (INIS)

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

1996-01-01

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

Structure and creep of Russian reactor steels with a BCC structure

Science.gov (United States)

Sagaradze, V. V.; Kochetkova, T. N.; Kataeva, N. V.; Kozlov, K. A.; Zavalishin, V. A.; Vil'danova, N. F.; Ageev, V. S.; Leont'eva-Smirnova, M. V.; Nikitina, A. A.

2017-05-01

The structural phase transformations have been revealed and the characteristics of the creep and long-term strength at 650, 670, and 700°C and 60-140 MPa have been determined in six Russian reactor steels with a bcc structure after quenching and high-temperature tempering. Creep tests were carried out using specially designed longitudinal and transverse microsamples, which were fabricated from the shells of the fuel elements used in the BN-600 fast neutron reactor. It has been found that the creep rate of the reactor bcc steels is determined by the stability of the lath martensitic and ferritic structures in relation to the diffusion processes of recovery and recrystallization. The highest-temperature oxide-free steel contains the maximum amount of the refractory elements and carbides. The steel strengthened by the thermally stable Y-Ti nanooxides has a record high-temperature strength. The creep rate at 700°C and 100 MPa in the samples of this steel is lower by an order of magnitude and the time to fracture is 100 times greater than that in the oxide-free reactor steels.

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

International Nuclear Information System (INIS)

Hsu Wensheng; Chen Hungpei; Hung Zhenyu; Lin Huichen

2014-01-01

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

Non-intuitive bubble effects in reactor and containment technology

International Nuclear Information System (INIS)

Moody, F.J.

1991-01-01

Most people know a lot about bubbles, including how they rise in liquids and the way they appear when the cap is removed from a bottle of carbonated beverage. A lot of bubble knowledge is obtained from bubbling air through water in aquariums to keep the fish alive and happy, or watching scuba divers feed the sharks in large glass tanks at the local zoo. But innocent bubbles can be sources of structural loadings and sometimes destructive fluid behavior. In fact, there are many non-intuitive effects associated with bubbles which have been discovered by experiments and analyses. It has been necessary to design various reactor and containment components in the nuclear energy industry to accommodate the fact that bubbles can expand like compressed springs, or oscillate, or collapse abruptly, and create structural loads. This paper describes several important phenomena associated with bubble action in nuclear reactor and containment systems and the associated loads exerted. An awareness of these effects can help to avoid unwelcome surprises in general thermal-hydraulic applications when a system is disturbed by bubble behavior. Major topics discussed include expanding and collapsing submerged bubbles, steam chugging and ringout, bubble shattering, surprising hot bubble action in a saturated pool, bubble effects on fluid-structure-interaction, waterhammer from collapsing bubble in pipes, and vapor bubble effects on sound speed in saturated mixtures

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

Structural design of SBWR reactor building complex using microcomputers

International Nuclear Information System (INIS)

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

1993-01-01

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

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.)

Assessment of corrosion and fatigue damage to light water reactor metal containments

International Nuclear Information System (INIS)

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

1991-01-01

This paper presents a generic procedure for estimating aging damage, evaluating structural integrity, and identifying mitigation activities for safe operation of boiling water reactor (BWR) Mark I metal containments and ice-condenser type pressurized water reactor (PWR) cylindrical metal containments. The mechanisms of concern that can cause aging damage to these two types of containments are corrosion and fatigue. Assessment of fatigue damage to bellows is also described. Assessment of corrosion and fatigue damage described in this paper include: containment design features that are relevant to aging assessment, several corrosion and fatigue mechanisms, inspection of corrosion and fatigue damage, and mitigation of damage caused by these mechanisms. In addition, synergistic interaction between corrosion and fatigue is considered. Possible actions for mitigating aging include enhanced inspection methods, maintenance activities based on operating experience, and supplementary surveillance programs. Field experience related to aging of metal containments is reviewed. Finally, conclusions and recommendations are presented

Reactor core structure

International Nuclear Information System (INIS)

Higashinakagawa, Emiko; Sato, Kanemitsu.

1992-01-01

Taking notice on the fact that Fe based alloys and Ni based alloys are corrosion resistant in a special atmosphere of a nuclear reactor, Fe or Ni based alloys are applied to reactor core structural components such as fuel cladding tubes, fuel channels, spacers, etc. On the other hand, the neutron absorption cross section of zirconium is 0.18 barn while that of iron is 2.52 barn and that of nickel is 4.6 barn, which amounts to 14 to 25 times compared with that of zirconium. Accordingly, if the reactor core structural components are constituted by the Fe or Ni based alloys, neutron economy is lowered. Since it is desirable that neutrons contribute to uranium fission with least absorption to the reactor core structural components, the reactor core structural components are constituted with the Fe or Ni based alloys of good corrosion resistance only at a portion in contact with reactor water, that is, at a surface portion, while the main body is constituted with zircalloy in the present invention. Accordingly, corrosion resistnace can be kept while keeping small neutron absorption cross section. (T.M.)

The CANDU 3 containment structure

International Nuclear Information System (INIS)

1994-01-01

The design of the CANDU 3 nuclear power plant is being developed by AECL CANDU's Saskatchewan office. There are 24 CANDU nuclear power units operating in Canada and abroad and eight units are under construction is Romania and South Korea. The design of the CANDU 3 plant has evolved on the basis of the proven CANDU design. The experiences gained during construction, commissioning and operation of the existing CANDU plants are considered in the design. Many technological enhancements have been implemented in the design processes in all areas. The object has been to develop an improved reactor design that is suitable for the current and the future markets worldwide. Throughout the design phase of CANDU 3, emphasis has been placed in reducing the cost and construction schedule of the plant. This has been achieved by implementing design improvements and using new construction techniques. Appropriate changes and improvements to the design to suit new requirements are also adopted. In CANDU plants, the containment structure acts as an ultimate barrier against the leakage of radioactive substances during normal operations and postulated accident conditions. The concept of the structural design of the containment structure has been examined in considerable detail. This has resulted in development of a new conceptual design for the containment structure for CANDU 3. This paper deals with this new design of the containment structure

Systems and methods for enhancing isolation of high-temperature reactor containments

Energy Technology Data Exchange (ETDEWEB)

Peterson, Per F.

2017-09-26

A high-temperature containment-isolation system for transferring heat from a nuclear reactor containment to a high-pressure heat exchanger is presented. The system uses a high-temperature, low-volatility liquid coolant such as a molten salt or a liquid metal, where the coolant flow path provides liquid free surfaces a short distance from the containment penetrations for the reactor hot-leg and the cold-leg, where these liquid free surfaces have a cover gas maintained at a nearly constant pressure and thus prevent high-pressures from being transmitted into the reactor containment, and where the reactor vessel is suspended within a reactor cavity with a plurality of refractory insulator blocks disposed between an actively cooled inner cavity liner and the reactor vessel.

Structural dynamics in fast reactor accident analysis

International Nuclear Information System (INIS)

Fistedis, S.H.

1975-01-01

Analyses and codes are under development combining the hydrodynamics and solid mechanics (and more recently the bubble dynamics) phenomena to gage the stresses, strains, and deformations of important primary components, as well as the overall adequacy of primary and secondary containments. An arbitrary partition of the structural components treated evolves into (1) a core mechanics effort; and (2) a primary system and containment program. The primary system and containment program treats the structural response of components beyond the core, starting with the core barrel. Combined hydrodynamics-solid mechanics codes provide transient stresses and strains and final deformations for components such as the reactor vessel, reactor cover, cover holddown bolts, as well as the pulses for which the primary piping system is to be analyzed. Both, Lagrangian and Eulerian two-dimensional codes are under development, which provide greater accuracy and longer durations for the treatment of HCDA. The codes are being augmented with bubble migration capability pertaining to the latter stages of the HCDA, after slug impact. Recent developments involve the adaptation of the 2-D Eulerian primary system code to the 2-D elastic-plastic treatment of primary piping. Pulses are provided at the vessel-primary piping interfaces of the inlet and outlet nozzles, calculation includes the elbows and pressure drops along the components of the primary piping system. Recent improvements to the primary containment codes include introduction of bending strength in materials, Langrangian mesh regularization techniques, and treatment of energy absorbing materials for the slug impact. Another development involves the combination of a 2-D finite element code for the reactor cover with the hydrodynamic containment code

Drop analysis for structural integrity evaluation of KJRR fuel transport container

International Nuclear Information System (INIS)

Yang, Yun Young; Lim, Jong Min; Choi, Woo Seok; Lee, Ju Chan

2016-01-01

A fuel transport container for KiJang Research Reactor(KJRR) has been developed to transport fresh fuel assemblies and fission molly targets which are used for a research reactor built in Kijang. The KJRR fuel transport container is a type-A(F) container, which is defined in domestic and foreign regulations of a radioactive substance container. According to Nuclear Safety and Security Commission's notification, the container should meet the accident conditions defined in IAEA safety Standard Series, US NRC and etc. In this study, a structural integrity of the KJRR fuel transport container is evaluated by conducting computational analyses of 9-meter free drop and 1 meter puncture. It is confirmed that structural integrity of the KJRR fuel transport container can be maintained in the transportation accident condition. Hereafter, when the test model is produced, a safety test will be conducted and its result will be compared with the result of drop and puncture analyses.

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

Corrosion failure of a BWR embedded reactor containment liner

International Nuclear Information System (INIS)

Wegemar, B.

2006-01-01

Following sixteen fuel cycles, leakage through a BWR embedded reactor containment liner (carbon steel) was discovered. Leakage was located at a penetration for electrical conductors as a result of penetrating corrosion attack. During construction, porous cement structures and air pockets/cavities were formed due to erroneous injection of grout. Corrosion attacks on the CS steel liner were located at the relatively small, active surfaces in contact with the porous cement structure. The corrosion mechanism was supposed to be anodic dissolution of the steel liner in areas with insufficient passivation. The penetrations were restored according to original design requirements. (author)

Proceedings of 18th international conference on structural mechanics in reactor technology

International Nuclear Information System (INIS)

2005-07-01

The 18th International Conference on Structural Mechanics in Reactor Technology was held on August 7-12, 2005 in Beijing, China, and Sponsored by International Association for Structural Mechanics in Reactor Technology, Chinese Nuclear Society, Chinese Society of Theoretical and Applied Mechanics, and Tsinghua University. 486 abstracts are Collected. The contents includes: opening, plenary and keynote presentations; computational mechanics; fuel and core structures; aging, life extension, and license renewal; design methods and rules for components; fracture mechanics; concrete material, containment and other structures; analysis and design for dynamic and extreme loads; seismic analysis, design and qualification; structural reliability and probabilistic safety assessment (PSA); operation, inspection and maintenance; severe accident management and structural evaluation; advanced reactors and generation IV reactors; decommissioning of nuclear facilities and waste management.

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)

Overpressurization performance of containment structures

International Nuclear Information System (INIS)

Barr, P.; Bleackley, M.; Harrop, L.P.; Hargreaves, J.; Jowett, J.; Phillips, D.W.

1987-01-01

The containment building of a PWR is the outermost engineered barrier between the reactor and the environment. The most important element of such a containment system is the pressure boundary structure and its associated seals and penetrations. This containment structure is designed deterministically to withstand a number of loads and load combinations of which the dominant one is generally the internal pressure due to the double-ended guillotine break in one of the primary circuit loops. Typically, the design basis large LOCA produces a peak pressure increase in the region of 0.3 MPa in some 10 seconds and with a duration of up to a few tens of seconds. The assessment of overpressure performance of the containment structure is a key component of the PWR safety case, and is usually carried out by estimating a static factor of safety to some failure limit state. These estimates can be made using simple force-balance calculations or complicated finite element calculations, and both approaches have merit. In this paper we examine these approaches and discuss their value in estimating failure pressures and failure modes for a variety of internal pressurization transients. This discussion covers both general design and risk considerations and is illustrated by numerical examples taken from previous and on-going analysis

Permanent cavity seal ring for a nuclear reactor containment arrangement

International Nuclear Information System (INIS)

Swidwa, K.J.; Salton, R.B.; Marshall, J.R.

1990-01-01

This patent describes a nuclear reactor containment arrangement. It comprises: a reactor pressure vessel which thermally expands and contracts during cyclic operation of the reactor, the vessel having a peripheral wall and a horizontally outwardly extending flange thereon; a containment wall having a shelf, the wall spaced from and surrounding the peripheral wall of the reactor pressure vessel defining an annular expansion gap therebetween, and an annular ring seal extending across the annular expansion gap to provide a water-tight seal therebetween

Protective guide structure for reactor control rod

International Nuclear Information System (INIS)

Ban, Minoru; Umeda, Kenji; Kubo, Noboru; Ito, Tomohiro.

1996-01-01

The present invention provides an improved protective guide structure for control rods, which does not cause swirling of coolants and resonance even though a slit is formed on a protective tube which surrounds a control rod element in a PWR type reactor. Namely, a reactor control rod is constituted with elongated control elements collectively bundled in the form of a cluster. The protective guide structure protectively guides the collected constituent at the upper portion of a reactor container. The protective structure comprises a plurality of protective tubes each having a C-shaped cross section disposed in parallel for receiving control rod elements individually in which the corners of the opening of the cross section of the protective tube are chamfered to an appropriate configuration. With such a constitution, even if coolant flows in a circumferential direction along the protective tubes surrounding the control rod elements, no shearing stream is caused to the coolants flow since the corners of the cross sectional opening (slit) of the tube are chamfered. Accordingly, occurrence of swirlings can be suppressed. (I.S.)

Containment and surveillance techniques at power reactors

International Nuclear Information System (INIS)

Stirling, A.J.

1982-01-01

This session will provide participants with an understanding of the functions of safeguards equipment at power reactors, including equipment for fuel accounting, video and film surveillance, diversion monitoring, and containment and surveillance of irradiated fuel in storage. In addition, some appreciation of the impact that reactor safeguards have on the plant operator will be gained. From this, participants will be able to ensure that a reactor safeguards system meets their nation's international and national nonproliferation objectives with a minimum of interference to plant operations

Variable stiffness lattice support system for a condenser type nuclear reactor containment

International Nuclear Information System (INIS)

George, J.A.; Sutherland, J.D.

1979-01-01

A support structure for the lattice supporting a fusible material in the annular condenser region of a nuclear reactor containment, the flexibility of which structure can be selectively adjusted in accordance with seismic or other loading requirements. The lattice is affixed to a flexible member in a manner which allows relative movement between the two components. The flexible member is affixed to a rigid support member in a manner which selectively adjusts the resiliency of the flexible member. The support member is rigidly affixed to a wall of the containment annulus, and can also be utilized to support cooling ducts. 6 claims

Method of constructing lower dry well access tunnel for nuclear reactor container

International Nuclear Information System (INIS)

Kume, Tadashi; Furukawa, Hedeyasu.

1993-01-01

The method of the present invention facilitates construction of a lower dry well access tunnel for a nuclear reactor container. The lower dry well access tunnel is constructed across the reactor container and the reactor main body foundation. In this case, the lower dry well access tunnel is divided into three sections, i.e., axial end portions and a central portion. At first, each of the end portions is attached to the walls of the reactor container and the reactor main body foundation respectively. Subsequently, the central portion is attached to each of the end portions. An adjusting margin is previously provided to the central portion upon manufacturing each of the sections for adjusting deviation from a nominal size upon construction. In such a construction method, it is possible to eliminate interference accident during construction between the end portions of the lower dry well access tunnel and the reactor container and the reactor main body foundation, to facilitate the construction. Further, this facilitates the fabricating operation for dimensional alignment between the lower dry well access tunnel, and the reactor container and the reactor main body foundation. (I.S.)

Tritium containment of controlled thermonuclear fusion reactor

International Nuclear Information System (INIS)

Tanaka, Yoshihisa; Tsukumo, Kiyohiko; Suzuki, Tatsushi

1979-01-01

It is well known that tritium is used as the fuel for nuclear fusion reactors. The neutrons produced by the nuclear fusion reaction of deuterium and tritium react with lithium in blankets, and tritium is produced. The blankets reproduce the tritium consumed in the D-T reaction. Tritium circulates through the main cooling system and the fuel supply and evacuation system, and is accumulated. Tritium is a radioactive substance emitting β-ray with 12.6 year half-life, and harmful to human bodies. It is an isotope of hydrogen, and apt to diffuse and leak. Especially at high temperature, it permeates through materials, therefore it is important to evaluate the release of tritium into environment, to treat leaked tritium to reduce its release, and to select the method of containing tritium. The permeability of tritium and its solubility in structural materials are discussed. The typical blanket-cooling systems of nuclear fusion reactors are shown, and the tungsten coating of steam generator tubes and tritium recovery system are adopted for reducing tritium leak. In case of the Tokamak type reactor of JAERI, the tritium recovery system is installed, in which the tritium gas produced in blankets is converted to tritium steam with a Pd-Pt catalytic oxidation tower, and it is dehydrated and eliminated with a molecular sieve tower, then purified and recovered. (Kako, I.)

Nuclear reactor container

International Nuclear Information System (INIS)

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

1994-01-01

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

Reactor coolant system and containment aqueous chemistry

International Nuclear Information System (INIS)

Torgerson, D.F.

1986-01-01

Fission products released from fuel during reactor accidents can be subject to a variety of environments that will affect their ultimate behavior. In the reactor coolant system (RCS), for example, neutral or reducing steam conditions, radiation, and surfaces could all have an effect on fission product retention and chemistry. Furthermore, if water is encountered in the RCS, the high temperature aqueous chemistry of fission products must be assessed to determine the quantity and chemical form of fission products released to the containment building. In the containment building, aqueous chemistry will determine the longer-term release of volatile fission products to the containment atmosphere. Over the past few years, the principles of physical chemistry have been rigorously applied to the various chemical conditions described above. This paper reviews the current state of knowledge and discusses the future directions of chemistry research relating to the behavior of fission products in the RCS and containment

Common floor system vertical earthquake-proof structure for reactor equipment

International Nuclear Information System (INIS)

Morishita, Masaki.

1996-01-01

In an LMFBR type reactor, a reactor container, a recycling pump and a heat exchanger are disposed on a common floor. Vertical earthquake-proof devices which can be stretched only in vertical direction formed by laminating large-sized bellevilles are disposed on a concrete wall at the circumference of each of reactor equipments. A common floor is placed on all of the vertical earthquake-proof devices to support the entire earthquake-proof structure simultaneously. If each of reactor equipments is loaded on the common floor and the common floor is entirely supported against earthquakes altogether, since the movement of each of the reactor equipments loaded on the common floor is identical, relative dislocation is not exerted on the main pipelines which connect the equipments. In addition, since the entire earthquake structure has a flat common floor and each of the reactor equipments is suspended to minimize the distance between a gravitational center and a support point, locking vibration is less caused to the horizontal earthquake. (N.H.)

Structural integrity evaluation of PWR nuclear reactor pressure vessels

International Nuclear Information System (INIS)

Cruz, Julio R.B.; Mattar Neto, Miguel

1999-01-01

The reactor pressure vessel (RPV) is the most important structural component of a PWR nuclear power plant. It contains the reactor core and is the main component of the primary system pressure boundary, the system responsible for removing the heat generated by the nuclear reactions. It is considered not replaceable and, therefore, its lifetime is a key element to define the plant life as a whole. Three critical issues related to the reliability of the RPV structural integrity come out by reason of the radiation damage imposed to the vessel material during operation. These issues concern the definition of pressure versus temperature limits for reactor heatup and cooldown, pressurized thermal shock evaluation and assessment of reactor vessels with low upper shelf Charpy impact energy levels. This work aims to present the major aspects related to these topics. The requirements for preventing fracture of the RPV are reviewed as well as the available technology for assessing the safety margins. For each mentioned problem, the several steps for structural integrity evaluation are described and the analysis methods are discussed. (author)

Not a mystery. Inner containment of the pressurized water reactor (EPR trademark type)

Energy Technology Data Exchange (ETDEWEB)

Ostermann, Dirk; Wienand, Burkhard; Krumb, Christian [AREVA NP GmbH (Germany)

2012-11-01

The containment of the advanced pressurized water reactor EPR trademark type is developed on the basis of the French nuclear power plant operational experience and consists of - The reinforced outer containment structure, designed to withstand external hazards (e.g. APC), - The pre-stressed inner containment structure, designed to bear the loads resulting from internal hazards (LOCA), - The steel liner, designed to provide leak tightness resulting from internal hazards. The main advantage of the pre-stressed inner containment design is that the structure remains in linear-elastic behavior during the whole life-time. Even in case of postulated design accidents (LOCA) concrete tensile strains are strongly limited. Due to pre-stressing the concrete structure remains practically free of cracks. Due to pre-stressing the leak tightness ensuring steel liner, embedded into the inner concrete shell, is exposed to more favorable compression loads. In addition to detailed calculations several test programs have been performed to verify and confirm the predicted behavior in normal operation and in accident condition. (orig.)

Proceedings of the twenty-fourth water reactor safety information meeting. Volume 1: Plenary session; High burnup fuel; Containment and structural aging

International Nuclear Information System (INIS)

Monteleone, S.

1997-01-01

This three-volume report contains papers presented at the Twenty-Fourth Water Reactor Safety Information Meeting held at the Bethesda Marriott Hotel, Bethesda, maryland, October 21--23, 1996. The papers are printed in the order of their presentation in each session and describe progress and results of programs in nuclear safety research conducted in this country and abroad. Foreign participation in the meeting included papers presented by researchers from Czech Republic, Finland, France, Japan, Norway, Russia and United Kingdom. This first volume is divided into 3 sections: plenary session; high burnup fuel; and containment and structural aging. Selected papers are indexed separately for inclusion in the Energy Science and Technology Database

Proceedings of the twenty-fourth water reactor safety information meeting. Volume 1: Plenary session; High burnup fuel; Containment and structural aging

Energy Technology Data Exchange (ETDEWEB)

Monteleone, S. [comp.] [Brookhaven National Lab., Upton, NY (United States)

1997-01-01

This three-volume report contains papers presented at the Twenty-Fourth Water Reactor Safety Information Meeting held at the Bethesda Marriott Hotel, Bethesda, maryland, October 21--23, 1996. The papers are printed in the order of their presentation in each session and describe progress and results of programs in nuclear safety research conducted in this country and abroad. Foreign participation in the meeting included papers presented by researchers from Czech Republic, Finland, France, Japan, Norway, Russia and United Kingdom. This first volume is divided into 3 sections: plenary session; high burnup fuel; and containment and structural aging. Selected papers are indexed separately for inclusion in the Energy Science and Technology Database.

Residual stress improving method for reactor structural component and residual stress improving device therefor

Energy Technology Data Exchange (ETDEWEB)

Enomoto, Kunio; Otaka, Masahiro; Kurosawa, Koichi; Saito, Hideyo; Tsujimura, Hiroshi; Tamai, Yasukata; Urashiro, Keiichi; Mochizuki, Masato

1996-09-03

The present invention is applied to a BWR type reactor, in which a high speed jetting flow incorporating cavities is collided against the surface of reactor structural components to form residual compression stresses on the surface layer of the reactor structural components thereby improving the stresses on the surface. Namely, a water jetting means is inserted into the reactor container filled with reactor water. Purified water is pressurized by a pump and introduced to the water jetting means. The purified water jetted from the water jetting means and entraining cavities is abutted against the surface of the reactor structural components. With such procedures, since the purified water is introduced to the water jetting means by the pump, the pump is free from contamination of radioactive materials. As a result, maintenance and inspection for the pump can be facilitated. Further, since the purified water injection flow entraining cavities is abutted against the surface of the reactor structural components being in contact with reactor water, residual compression stresses are exerted on the surface of the reactor structural components. As a result, occurrence of stress corrosion crackings of reactor structural components is suppressed. (I.S.)

Residual stress improving method for reactor structural component and residual stress improving device therefor

International Nuclear Information System (INIS)

Enomoto, Kunio; Otaka, Masahiro; Kurosawa, Koichi; Saito, Hideyo; Tsujimura, Hiroshi; Tamai, Yasukata; Urashiro, Keiichi; Mochizuki, Masato.

1996-01-01

The present invention is applied to a BWR type reactor, in which a high speed jetting flow incorporating cavities is collided against the surface of reactor structural components to form residual compression stresses on the surface layer of the reactor structural components thereby improving the stresses on the surface. Namely, a water jetting means is inserted into the reactor container filled with reactor water. Purified water is pressurized by a pump and introduced to the water jetting means. The purified water jetted from the water jetting means and entraining cavities is abutted against the surface of the reactor structural components. With such procedures, since the purified water is introduced to the water jetting means by the pump, the pump is free from contamination of radioactive materials. As a result, maintenance and inspection for the pump can be facilitated. Further, since the purified water injection flow entraining cavities is abutted against the surface of the reactor structural components being in contact with reactor water, residual compression stresses are exerted on the surface of the reactor structural components. As a result, occurrence of stress corrosion crackings of reactor structural components is suppressed. (I.S.)

Soil-structure interaction effects on the reliability evaluation of reactor containments

International Nuclear Information System (INIS)

Pires, J.; Hwang, H.; Reich, M.

1986-01-01

The probability-based method for the seismic reliability assessment of nuclear structures, which has been developed at Brookhaven National Laboratory (BNL), is extended to include the effects of soil-structure interaction. A reinforced concrete containment building is analyzed in order to examine soil-structure interaction effects on: (1) structural fragilities; (2) floor response spectra statistics; and (3) correlation coefficients for total acceleration responses at specified structural locations

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

International Nuclear Information System (INIS)

1997-08-01

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

NucleDyne's passive containment system

International Nuclear Information System (INIS)

Falls, O.B. Jr.; Kleimola, F.W.

1987-01-01

A simple definition of the passive containment system is that it is a total safeguards system for light water reactors designed to prevent and contain any accidental release of radioactivity. Its passive features utilize the natural laws of physics and thermodynamics. The system encompasses three basic containments constructed as one integrated structure on the reactor building foundation. The primary containment encloses the reactor pressure vessel and coolant system and passive engineered safety systems and components. Auxiliary containment enclosures house auxiliary systems and components. Secondary containment (the reactor building), housing the primary and auxiliary containment structures, provides a second containment barrier as added defense-in-depth against leakage of radioactivity for all accidents assumed by the industry. The generic features of the passive containment system are applicable to both the boiling water reactors and the pressurized water reactors as standardized features for all power ranges. These features provide for a zero source term, the industry's ultimate safety goal. This paper relates to a four-loop pressurized water reactor

Nuclear reactor container

International Nuclear Information System (INIS)

Shioiri, Akio.

1992-01-01

In a nuclear reactor container, a vent tube communication port is disposed to a pressure suppression pool at a position higher than the pool water therein for communication with an upper dry well, and the upper end opening of a dry well communication pipe is disposed at a position higher than the communication port. When condensate return pipeline is ruptured in the upper dry well, water in a water source pool is injected to the pressure vessel and partially discharged out of the ruptured port and a depressurization valve connected to the pressure vessel to the inside of the upper dry well. The discharged water stays in the upper dry well and, when the water level reaches the height of the vent tube communication port, it flows into the pressure suppression pool. Even in a state that the entire amount of water in the water source pool is supplied, since water does not reach the upper opening port of the dry well communication pipe, water does not flow into a lower dry well. Accordingly, the motor of a control rod drives disposed in the lower dry well can be prevented from submerging. The reactor core can be cooled more reliably, to improve the reliability of the pressure suppression function. (N.H.)

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

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)

Proof testing of CANDU concrete containment structures

International Nuclear Information System (INIS)

Pandey, M.D.

1996-05-01

Prior to commissioning of a CANDU reactor, a proof pressure test is required to demonstrate the structural integrity of the containment envelope. The test pressure specified by AECB Regulatory Document R-7 (1991) was selected without a rigorous consideration of uncertainties associated with estimates of accident pressure and conatinment resistance. This study was undertaken to develop a reliability-based philosophy for defining proof testing requirements that are consistent with the current limit states design code for concrete containments (CSA N287.3).It was shown that the upodated probability of failure after a successful test is always less than the original estimate

Reactor container

International Nuclear Information System (INIS)

Abe, Yoshihito; Sano, Tamotsu; Ueda, Sabuo; Tanaka, Kazuhisa.

1987-01-01

Purpose: To improve the liquid surface disturbance in LMFBR type reactors. Constitution: A horizontal flow suppressing mechanism mainly comprising vertical members is suspended near the free liquid surface of coolants in the upper plenum. The horizontal flow of coolants near the free liquid surface is reduced by the suppressing mechanism to effectively reduce the surface disturbance. The reduction in the liquid surface disturbance further prevails to the entire surface region with no particular vertical variations to the free liquid surface to remarkably improve the preventive performance for the liquid surface disturbance. Accordingly, it is also possible to attain the advantageous effects such as prevention for the thermal fatigue in reactor vessel walls, reactor upper mechanisms, etc. and prevention of burning damage to the reactor core due to the reduction of envolved Ar gas. (Kamimura, M.)

Behaviour of prestressed concrete containment structures

International Nuclear Information System (INIS)

MacGregor, J.G.; Murray, D.W.; Simmonds, S.H.

1980-05-01

The most significant finds from a study to assess the response of prestressed concrete secondary containment structures for nuclear reactors under the influence of high internal overpressures are presented. A method of analysis is described for determining the strains and deflections including effects of inelastic behaviour at various points in the structure resulting from increasing internal pressures. Experimentally derived relationships between the strains and crack spacing, crack width and leakage rate are given. These procedures were applied to the Gentilly-2 containment building to obtain the following results: (1) The first through-the-wall cracks would occur in the dome at 48 psi or 2.3 times the proof test pressure. (2) At this pressure leakage would begin and would increase exponentially as the pressure increases such that at 93% of the predicted failure load the calculated leakage rate would be approximately equal to the volume of the containment each second. (3) Assuming the pressurizing medium could be supplied sufficiently rapidly, failure would occur due to rupture of the horizontal tendons at approximately 77 psi. (author)

Development of the containment transient analysis code for the passive reactor

Energy Technology Data Exchange (ETDEWEB)

Hwang, Young Dong; Kim, Young In; Bae, Yoon Young; Chang, Moon Hi [Korea Atomic Energy Research Institute, Taejon (Korea)

1998-05-01

This study was performed to develop the analysis tools for the passively cooled steel containment and to construct the integrated code system which can analyze a thermal hydraulic behavior of the containment and reactor system during a loss of coolant accident. The computer code CONTEMPT4/MOD5/PCCS was developed by incorporating the passive containment cooling models to the containment pressure and temperature transient analysis computer code CONTEMPT4/MOD5. The integrated reactor thermal hydraulic analysis code system for passive reactor was constructed by coupling the best estimate thermal hydraulic system analysis code RELAP5/MOD3 and CONTEMPT4/MOD5/PCCS through the process control method. In addition, to evaluate the applicability of the code the CONTEMPT4/MOD5/PCCS was applied to the SMART(System-Integrated Modular Advanced Reactor). The pressure and temperature transient following the small break LOCA of SMART was analysed by modeling the safeguard vessel using both the newly added passive containment cooling model and existing pool model. (author). 16 refs., 22 figs., 7 tabs.

Models of iodine behavior in reactor containments

Energy Technology Data Exchange (ETDEWEB)

Weber, C.F.; Beahm, E.C.; Kress, T.S.

1992-10-01

Models are developed for many phenomena of interest concerning iodine behavior in reactor containments during severe accidents. Processes include speciation in both gas and liquid phases, reactions with surfaces, airborne aerosols, and other materials, and gas-liquid interface behavior. Although some models are largely empirical formulations, every effort has been made to construct mechanistic and rigorous descriptions of relevant chemical processes. All are based on actual experimental data generated at the Oak Ridge National Laboratory (ORNL) or elsewhere, and, hence, considerable data evaluation and parameter estimation are contained in this study. No application or encoding is attempted, but each model is stated in terms of rate processes, with the intention of allowing mechanistic simulation. Taken together, this collection of models represents a best estimate iodine behavior and transport in reactor accidents.

Models of iodine behavior in reactor containments

International Nuclear Information System (INIS)

Weber, C.F.; Beahm, E.C.; Kress, T.S.

1992-10-01

Models are developed for many phenomena of interest concerning iodine behavior in reactor containments during severe accidents. Processes include speciation in both gas and liquid phases, reactions with surfaces, airborne aerosols, and other materials, and gas-liquid interface behavior. Although some models are largely empirical formulations, every effort has been made to construct mechanistic and rigorous descriptions of relevant chemical processes. All are based on actual experimental data generated at the Oak Ridge National Laboratory (ORNL) or elsewhere, and, hence, considerable data evaluation and parameter estimation are contained in this study. No application or encoding is attempted, but each model is stated in terms of rate processes, with the intention of allowing mechanistic simulation. Taken together, this collection of models represents a best estimate iodine behavior and transport in reactor accidents

Methodology for predicting ultimate pressure capacity of the ACR-1000 containment structure

International Nuclear Information System (INIS)

Saudy, A.M.; Awad, A.; Elgohary, M.

2006-01-01

The Advanced CANDU Reactor or the ACR-1000 is developed by Atomic Energy of Canada Limited (AECL) to be the next step in the evolution of the CANDU product line. It is based on the proven CANDU technology and incorporates advanced design technologies. The ACR containment structure is an essential element of the overall defense in depth approach to reactor safety, and is a physical barrier against the release of radioactive material to the environment. Therefore, it is important to provide a robust design with an adequate margin of safety. One of the key design requirements of the ACR containment structure is to have an ultimate pressure capacity that is at least twice the design pressure Using standard design codes, the containment structure is expected to behave elastically at least up to 1.5 times the design pressure. Beyond this pressure level, the concrete containment structure with reinforcements and post-tension tendons behaves in a highly non-linear manner and exhibits a complex response when cracks initiate and propagate. To predict the structural non-linear responses, at least two critical features are involved. These are: the structural idealization by the geometry and material property models, and the adopted solution algorithm. Therefore, detailed idealization of the concrete structure is needed in order to accurately predict its ultimate pressure capacity. This paper summarizes the analysis methodology to be carried out to establish the ultimate pressure capacity of the ACR containment structure and to confirm that the structure meets the specified design requirements. (author)

Risk-informed assessment of degraded containment structures

International Nuclear Information System (INIS)

Spencer, B.W.; Kunsman, D.M.; Graves, H.L.

2003-01-01

As nuclear power plants age, a number of degradation mechanisms may begin to affect the ability of critical containment structures to prevent radiation release during a severe accident. A research program is underway to quantify the effects of various types of containment degradation in a risk-informed manner. In this paper, corrosion is assumed to occur in the liner of a reinforced concrete containment at a 'typical' U.S. pressurized water reactor nuclear power plant, and its effect is investigated. Latin hypercube sampling is used in conjunction with finite element models of a typical steel-lined reinforced concrete containment to generate overpressurization fragilities of the containment with and without corrosion. An existing probabilistic risk assessment model of the plant is then used with these fragilities to determine the increase in risk caused by the corrosion. (author)

EDF reactor building containment: Monitoring of the pre-stressed concrete structure

International Nuclear Information System (INIS)

Badez, N.

2009-01-01

The concrete containments of the EDF PWR are pre-stressed, and are monitored to observe the ageing effects on the structure, in particular the evolutions of creep, shrinkage, pre-stress loss, and air leakage tightness. Monitoring devices are installed during construction period, and measurements are checked, stored on a data base, and analysed during all the plant operating life time. The topic of the presentation is to present each part of the EDF monitoring organisation. A continuous monitoring makes it possible to produce periodical comprehensive reports about the mechanical analysis of the structure, the strain stabilisation,... Periodical tests (each 10 years) are planned. They consist to submit the containment to an internal air pressure at the accidental pressure level. The monitoring system gives the strain values in order to check their linearity and reversibility with decreasing pressure. At the same time, the containment tightness is checked with a specific instrumentation to verify that leak rate is lower than the required level. A general view of instrumentation implemented on the containment (sensors, data acquisition), and a data analysis are presented

The economic potential of a cassette-type-reactor-installed nuclear ice-breaking container ship

International Nuclear Information System (INIS)

Kondo, K.; Takamasa, T.

2000-01-01

The design concept of the cassette-type-reactor MRX (Marine Reactor X), being under development in Japan for the nuclear ice-breaker container ship is described. The MRX reactor is the monoblock water-cooled and moderated reactor with passive cooling system of natural circulation. It is shown that application of the reactor being under consideration gives an opportunity to decrease greatly the difference in prices for similar nuclear and diesel ships. Economic estimations for applicability of the nuclear ice-breaker container ship with the MRX reactor in Arctics for transportation of standard containers TEU from Europe to Far East as compared with transportation of the same containers by diesel ships via Suets Canal are made [ru

Safety Research Experiment Facility Project. Conceptual design report. Volume IV. Reactor containment

International Nuclear Information System (INIS)

1975-12-01

The principal purpose of the SAREF Reactor Containment Building (RCB) is to prevent the uncontrolled release of radioactive materials to the atmosphere as a result of accidental occurrences inside the containment. The RCB houses numerous reactor systems and components including the Prestressed Concrete Reactor Vessel (PCRV). The design of the RCB is of reinforced concrete (steel-lined). The containment building is embedded nearly 100 feet in lava rock. It has therefore been necessary to independently formulate an appropriate and conservative design approach

Actions to reduce radioactive emissions: prevention of containment failure by flooding Containment and Reactor Cavity

International Nuclear Information System (INIS)

Fornos Herrando, J.

2013-01-01

The reactor cavity of Asco and Vandellos II is dry type, thus a severe accident leading to vessel failure might potentially end up resulting in the loss of containment integrity, depending on the viability to cool the molten core. Therefore, significant radioactive emissions could be released to outside. In the framework of Fukushima Stress Tests, ANAV has analyzed the convenience of carrying out different actions to prevent failure of the containment integrity in order to reduce radioactive emissions. The aim of this paper is to present and describe the main phenomenological aspects associated with two of these actions: containment flooding and reactor cavity flooding.

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

Science.gov (United States)

Zhou, Jianguo; Xu, Yaming; Zhang, Tao

2016-06-14

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

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

International Nuclear Information System (INIS)

Fontana, M.H.

1993-01-01

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

Filtered atmospheric venting of light water reactor containments

International Nuclear Information System (INIS)

Hedgran, A.; Ahlstroem, P.E.; Nilsson, L.; Persson, Aa.

1982-11-01

The aim of filtered venting is to improve the function of the reactor containment in connection with very severe accidents. By equipping the containment with a safety valve for pressure relief and allowing the released gases to pass through an effective filter, it should be possible to achieve a considerable protective effect. The work has involved detailed studies of the core meltdown sequence, how the molten core material runs out of the reactor vessel, what effect it has on concrete and other structures and how final cooling of the molten core material takes place. On the basis of previous Swedish studies, the project has chosen to study a filter concept that consists of a gravel bed of large volume. This filter plant shall not only retain the radioactive particles that escape from the containment through the vent line, but shall also condense the accompanying steam. After the government decided in 1981 that Barsebaeck was to be equipped with filtered venting and issued specifications regarding its performance, the project aimed at obtaining results that could be used to design and verify a plant for filtered venting at the Barsebaeck nuclear power station. As far as the other Swedish nuclear power plants at Oskarshamn, Ringhals and Forsmark are concerned, the results are only applicable to a limited extent. Additional studies are required for these nuclear power plants before the value of filtered venting can be assessed. Based on the results of experiments and analyses, the project has made a safety analysis with Barsebaeck as a reference plant in order to study how the introduction of filtered venting affects the safety level at a station. In summary, the venting function appears to entail a not insignificant reduction of risks for boiling water reactors of the Barsebaeck type. For a number of types of such very severe core accident cases, the filter design studied ensures a substantial reduction of the releases. However it has not been possible within the

Development of fast reactor metal fuels containing minor actinides

International Nuclear Information System (INIS)

Ohta, Hirokazu; Ogata, Takanari; Kurata, Masaki; Koyama, Tadafumi; Papaioannou, Dimitrios; Glatz, Jean-Paul; Rondinella, Vincenzo V.

2011-01-01

Fast reactor metal fuels containing minor actinides (MAs) Np, Am, and Cm and rare earths (REs) Y, Nd, Ce, and Gd are being developed by the Central Research Institute of Electric Power Industry (CRIEPI) in collaboration with the Institute for Transuranium Elements (ITU) in the METAPHIX project. The basic properties of U-Pu-Zr alloys containing MA (and RE) were characterized by performing ex-reactor experiments. On the basis of the results, test fuel pins including U-Pu-Zr-MA(-RE) alloy ingots in parts of the fuel stack were fabricated and irradiated up to a maximum burnup of ∼10 at% in the Phenix fast reactor (France). Nondestructive postirradiation tests confirmed that no significant damage to the fuel pins occurred. At present, detailed destructive postirradiation examinations are being carried out at ITU. (author)

Containment atmosphere cooling system for experimental fast reactor 'JOYO'

International Nuclear Information System (INIS)

Sasaki, Mikio; Hoshi, Akio; Sato, Morihiko; Takeuchi, Kaoru

1979-01-01

The experimental fast reactor ''JOYO'', the first sodium-cooled fast reactor in Japan, achieved the initially licensed full power operation (50 MW) in July 1978 and is now under steady operation. Toshiba has participated in the construction of this reactor as a leading manufacturer and supplied various systems. This article outlines the design philosophy, system concepts and the operating experience of the containment atmosphere cooling system which has many design interfaces throughout the whole plant and requires especially high reliability. The successful performance of this system during the reactor full-power operation owes to the spot cooling design philosophy and to the preoperational adjustment of heat load during the preheating period of reactor cooling system peculiar to FBR. (author)

Leak preventing method and device for penetration portion in reactor container

International Nuclear Information System (INIS)

Sato, Mitsuyoshi.

1995-01-01

A pressurized air is sent to a leak path caused in an electric wire penetration portion on the side of a reactor container to keep the pressure thereof higher than that at the inside of the reactor container, or aerosol is injected to close the leak path. Alternatively, a cooling tube is disposed for cooling a sleeve, or a resin charging device is disposed for injecting a thermosetting resin to close the leak path, and an external power operated compressor is disposed for sending a pressurized air to the resin charging device. Then, pressure is applied in the opposite direction from a pressurizing port so as to provide a higher pressure than the inside of the reactor container by the compressor upon occurrence of an accident. In addition, the sleeve is cooled so as to suppress the less leak path in the wire penetration portion. Further, the leak path is closed by the aerosol or the thermosetting resin, which is quite effective also in view of improvement of the reliability of the reactor container. (N.H.)

Effect of high temperature on integrity of concrete containment structures

International Nuclear Information System (INIS)

Bhat, P.D.

1986-01-01

The effect of high temperature on concrete material properties and structural behavior are studied in order to relate these effects to the performance of concrete containment structures. Salient data obtained from a test program undertaken to study the behavior of a restrained concrete structure under thermal gradient loads up to its ultimate limit are described. The preliminary results indicate that concrete material properties can be considered to remain unaltered up to temperatures of 100 0 C. The presence of thermal gradients did not significantly affect the structures ultimate mechanical load capacity. Relaxation of restraint forces due to creep was found to be an important factor. The test findings are compared with the observations made in available literature. The effect of test findings on the integrity analysis of a containment structure are discussed. The problem is studied from the viewpoint of a CANDU heavy water reactor containment

Nuclear reactor container

International Nuclear Information System (INIS)

Fukui, Tooru; Murase, Michio; Kataoka, Yoshiyuki; Hidaka, Masataka; Sumita, Isao; Tominaga, Kenji.

1992-01-01

In a nuclear reactor container, a chamber in communication with a wet well of a pressure suppression chamber is disposed and situated to such a position that the temperature is lower than a chamber containing pool water upon occurrence of loss of coolant accident. In addition, the inner surface of the pressure suppression chamber is constituted with steel walls in contact with pool water, and an outer circumferential pool is disposed at the outer circumferential surface thereof. Further, a circulation channel is disposed, and a water intake port is disposed at a position higher than an exit to the pool water, and a water discharge port is opened in the pool water at a position lower than the exit to the pool water. With such a constitution, the allowable temperature of the pressure suppression pool water can be elevated to a saturated steam temperature corresponding to the resistant pressure of the container, so that the temperature difference between the pressure suppression pool and the outer side thereof is increased by so much, to improve thermal radiation performance. Accordingly, it can be utilized as a pressure suppression means for a plant of greater power. Further, thermal conduction efficiency from the pool water region of the pressure suppression chamber to the outer circumferential pool water is improved, or thermal radiation area is enlarged due to the circulation channel, to improve the heat radiation performance. (N.H.)

Pressure suppression facility for reactor container

International Nuclear Information System (INIS)

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

1993-01-01

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

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.)

Optimal Protection of Reactor Hall Under Nuclear Fuel Container Drop Using Simulation Methods

Directory of Open Access Journals (Sweden)

Králik Juraj

2014-12-01

Full Text Available This paper presents of the optimal design of the damping devices cover of reactor hall under impact of nuclear fuel container drop of type TK C30. The finite element idealization of nuclear power plant structure is used in software ANSYS. The steel pipe damper system is proposed for dissipation of the kinetic energy of the container free fall in comparison with the experimental results. The probabilistic and sensitivity analysis of the damping devices was considered on the base of the simulation methods in program AntHill using the Monte Carlo method.

Failure/leakage predictions of concrete structures containing cracks

International Nuclear Information System (INIS)

Pan, Y.C.; Marchertas, A.H.; Kennedy, J.M.

1984-06-01

An approach is presented for studying the cracking and radioactive release of a reactor containment during severe accidents and extreme environments. The cracking of concrete is modeled as the blunt crack. The initiation and propagation of a crack are determined by using the maximum strength and the J-integral criteria. Furthermore, the extent of cracking is related to the leakage calculation by using a model developed by Rizkalla, Lau and Simmonds. Numerical examples are given for a three-point bending problem and a hypothetical case of a concrete containment structure subjected to high internal pressure during an accident

Study to evaluate the feasibility of constructing a retrofit containment for the 105-L reactor at the Savannah River Plant

International Nuclear Information System (INIS)

Quinn, R.D.

1989-01-01

This paper presents a summary of a study performed to determine the feasibility of constructing a retrofit containment dome meeting the requirements of the ASME Boiler and Pressure Vessel Code for nuclear containment vessels over the existing Savannah River 105-L reactor. Using existing large dome structures as a guide, design concepts were developed and analyses performed to evaluate the structural feasibility of containment dome structures. Construction schedules and costs were estimated to assess financial feasibility as well. It was concluded that such a retrofit containment dome was structurally feasible and within the capabilities of present day construction technology

Proceedings of the 17th international conference on structural mechanics in reactor technology

International Nuclear Information System (INIS)

2003-01-01

The conference was divided into the following divisions and subdivisions: DIVISION A: Plenary lectures and panel; DIVISION B: Computational mechanics (Structural and thermal analysis; High-non linear analysis, material behaviour; Vibration and fluid dynamics analysis); DIVISION C: Fuel and core structures (Fuel vibration and fretting; Fuel design and constitutive modelling; Fuel failure under operation and accident conditions; Fuel failure under operation and accident conditions; Components and material behaviour under irradiation; Integrity of fuel systems under transient conditions); DIVISION D: Aging, Life Extension and Licence Renewal (International Regulatory and Economic Perspectives; Utility perspectives, WWER technology; Fatigue, corrosion and crack issues; Component integrity; Aging assessment and monitoring; Containment and other structures); DIVISION F: Design methods and rules for components (International codes and standards; Tube, piping codes and standards; Analyses; Fatigue and life assessment; Creep; Bolted connections and gaskets); DIVISION G: Fracture mechanics (Reactor pressure vessel integrity; Dynamic loading; Fracture considerations for various applications; Failure assessment of Zr alloy; Pipe integrity; Integrity of welds; Failure of non-metallic materials; Leak before break (LBB); Corrosion aspects); DIVISION H: Concrete Containment and Other Structures (Concrete materials and performance; Tests of scale prestressed concrete containment vessel; Shear wall test and analysis; Structural analysis and containment design; Structural integrity and analysis); DIVISION J: Analysis and design for dynamic and extreme load (Vibration of shells and plates; Impact analysis; Piping vibration; Structural dynamics; Experimental and other topics); DIVISION K: Seismic analysis, design and qualification (General seismic issues; Ground motion and sitting; Soil-structure interaction; Seismic response of structures; Seismic re-evaluation; Seismic response and

Helium effect on mechanical property of fusion reactor structural materials

International Nuclear Information System (INIS)

Yamamoto, Norikazu; Chuto, Toshinori; Murase, Yoshiharu; Nakagawa, Johsei

2004-01-01

High-energy neutrons produced in fusion reactor core caused helium in the structural materials of fusion reactors, such as blankets. We injected alpha particles accelerated by the cyclotron to the samples of martensite steel (9Cr3WVTaB). Equivalent helium doses injected to the sample is estimated to be up to 300 ppm, which were estimated to be equivalent to helium accumulation after the 1-year reactor operation. Creep tests of the samples were made to investigate helium embrittlement. There were no appreciable changes in the relation between the stresses and the rupture time, the minimum creep rate and the applied stress. Grain boundary effect by helium was not observed in ruptured surfaces. Fatigue tests were made for SUS304 samples, which contain helium up to 150 ppm. After 0.05 Hz cyclic stress tests, it was shown that the fatigue lifetime (cycles to rupture and extension to failure) are 1/5 in 150 ppm helium samples compared with no helium samples. The experimental results suggest martensite steel is promising for structural materials of fusion reactors. (Y. Tanaka)

Seismic response analysis of reactor containment structures - axisymmetric model with modified ground motion

International Nuclear Information System (INIS)

Saha, S.; Dasgupta, A.; Basu, P.C.

1993-01-01

Seismic analysis of a Reactor Building is performed idealising the system as a beam model (BM) and also an Axi-symmetric model (ASM) and the results compared. In both the cases effect of Soil-Structure Interaction have been taken Into account. Since the lower boundary of the ASM was at a depth much lower than that of the BM, deconvolution of the specified Free-Field Motion (FFM) was necessary. The deconvolution has been performed using frequency domain approach. (author)

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

Heat removing device for nuclear reactor container facility

Energy Technology Data Exchange (ETDEWEB)

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

1994-09-30

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

Safety margins associated with containment structures under dynamic loading

International Nuclear Information System (INIS)

Lu, S.C.

1978-01-01

A technical basis for assessing the true safety margins of containment structures involved with MARK I boiling water reactor reevaluation activities is presented. It is based on the results of a plane-strain, large displacement, elasto-plastic, finite-element analysis of a thin cylindrical shell subjected to external and internal pressure pulses. An analytical procedure is presented for estimating the ultimate load capacity of the thin shell structure, and subsequently, for quantifying the design margins of safety for the type of loads under consideration. For defining failure of structures, a finite strain failure criterion is derived that accounts for multiaxiality effects

Compatibility of structural materials with fusion reactor coolant and breeder fluids

International Nuclear Information System (INIS)

DeVan, J.H.

1979-01-01

Fusion reactors are characterized by a lithium-containing blanket, a heat transfer medium that is integral with the blanket and first wall, and a heat engine that couples to the heat transfer medium. A variety of lithium-containing substances have been identified as potential blanket materials, including molten lithium metal, molten LiF-BeF 2 , Pb-Li alloys, and solid ceramic compounds such as Li 2 O. Potential heat transfer media include liquid lithium, liquid sodium, molten nitrates, water, and helium. Each of these coolants and blankets requires a particular set of chemical and mechanical properties with respect to the associated reactor and heat engine structural materials. This paper discusses the materials factors that underlie the selection of workable combinations of blankets and coolants. It also addresses the materials compatibility problems generic to those blanket-coolant combinations currently being considered in reactor design studies. (orig.)

Development of a double containment concept for the European pressurized water reactor

International Nuclear Information System (INIS)

Costaz, J.L.; Bonhomme, N.; L'Huby, Y.; Sidaner, J.F.

1994-01-01

This paper addresses the development of a double containment concept for the European Pressurized Water Reactor. Specification of containment leak tightness during severe hazards resulting from core melt scenarios is part of the safety goals defined for the EPR project. These safety goals include retention of molten core, mitigation of hydrogen deflagration or explosion risks and decay heat removal. The main new containment structural design loads which have been defined, including containment pressure and temperature conditions following possible postulated-core melt events, are recalled in the paper. The feasibility of a double containment with a prestressed concrete inner containment taking into account these new design loads but based upon experience gained within the well tested concept of concrete double wall containment used in 1400 MW nuclear power plants which have already been built in France, is presented. The main characteristics of such a prestressed inner containment are described. Limits and further possible optimization for even more severe design loads (including liner option) are indicated. Experimental works including a large scale mock up are already under way. (author). 2 refs., 4 figs

Aging of the containment pressure boundary in light-water reactor plants

International Nuclear Information System (INIS)

Naus, D.J.; Oland, C.B.; Ellingwood, B.R.

1997-01-01

Research is being conducted by the Oak Ridge National Laboratory to address aging of the containment pressure boundary in light-water reactor plants. The objectives of this work are to (1) identify the significant factors related to occurrence of corrosion, efficacy of inspection, and structural capacity reduction of steel containments and liners of concrete containments, and to make recommendations on use of risk models in regulatory decisions; (2) provide NRC reviewers a means of establishing current structural capacity margins for steel containments, and concrete containments as limited by liner integrity; and (3) provide recommendations, as appropriate, on information to be requested of licensees for guidance that could be utilized by NRC reviewers in assessing the seriousness of reported incidences of containment degradation. In meeting these objectives research is being conducted in two primary task areas - pressure boundary condition assessment and root-cause resolution practices, and reliability-based condition assessments. Under the first task area a degradation assessment methodology was developed for use in characterizing the in-service condition of metal and concrete containment pressure boundary components and quantifying the amount of damage that is present. An assessment of available destructive and nondestructive techniques for examining steel containments and liners is ongoing. Under the second task area quantitative structural reliability analysis methods are being developed for application to degraded metallic pressure boundaries to provide assurances that they will be able to withstand future extreme loads during the desired service period with a level of reliability that is sufficient for public safety. To date, mathematical models that describe time-dependent changes in steel due to aggressive environmental factors have been identified, and statistical data supporting their use in time-dependent reliability analysis have been summarized

CLASSIFICATION OF SYSTEMS FOR PASSIVE AFTERHEAT REMOVAL FROM REACTOR CONTAINMENT OF NUCLEAR POWER PLANT WITH WATER-COOLED POWER REACTOR

Directory of Open Access Journals (Sweden)

N. Khaled

2014-01-01

Full Text Available A classification on systems for passive afterheat removal from reactor containment has been developed in the paper.  The classification permits to make a detailed analysis of various concepts pertaining to systems for passive afterheat removal from reactor containment of new generation. The paper considers main classification features of the given systems.

Trends in large-scale testing of reactor structures

International Nuclear Information System (INIS)

Blejwas, T.E.

2003-01-01

Large-scale tests of reactor structures have been conducted at Sandia National Laboratories since the late 1970s. This paper describes a number of different large-scale impact tests, pressurization tests of models of containment structures, and thermal-pressure tests of models of reactor pressure vessels. The advantages of large-scale testing are evident, but cost, in particular limits its use. As computer models have grown in size, such as number of degrees of freedom, the advent of computer graphics has made possible very realistic representation of results - results that may not accurately represent reality. A necessary condition to avoiding this pitfall is the validation of the analytical methods and underlying physical representations. Ironically, the immensely larger computer models sometimes increase the need for large-scale testing, because the modeling is applied to increasing more complex structural systems and/or more complex physical phenomena. Unfortunately, the cost of large-scale tests is a disadvantage that will likely severely limit similar testing in the future. International collaborations may provide the best mechanism for funding future programs with large-scale tests. (author)

Bounding analysis of containment of high pressure melt ejection in advanced light water reactors

International Nuclear Information System (INIS)

Additon, S.L.; Fontana, M.H.; Carter, J.C.

1990-01-01

This paper reports on the loadings on containment due to direct containment heating (DCH) as a result of high pressure melt ejection (HPME) in advanced light water reactors (ALWR) which were estimated using conservative, bounding analyses. The purpose of the analyses was to scope the magnitude of the possible loadings and to indicate the performance needed from potential mitigation methods, such as a cavity configuration that limits energy transfer to the upper containment volume. Analyses were performed for three cases which examined the effect of availability of high pressure reactor coolant system water at the time of reactor vessel melt through and the effect of preflooding of the reactor cavity. The amount of core ejected from the vessel was varied from 100% to 0% for all cases. Results indicate that all amounts of core debris dispersal could be accommodated by the containment for the case where the reactor cavity was preflooded. For the worst case, all the energy from in-vessel hydrogen generation and combustion plus that from 45% of the entire molten core would be required to equilibrate with the containment upper volume in order to reach containment failure pressure

Dynamic analysis and structural design of underground nuclear reactor containments

International Nuclear Information System (INIS)

Kierans, T.W.; Reddy, D.V.

1975-01-01

All concept options are assumed to be similar in design criteria for structural competence to contain radioactivity and fuel heat and meet the functional, servicing, protective and aesthetic requirements. The choice of underground siting should be based on criteria developed from the sequential consideration of load-causing phenomena, concept and site characteristics. From the criteria, loads for a particular concept and site are calculated and the design formulated. (orig./ORU) [de

Nuclear reactor container

International Nuclear Information System (INIS)

Kawabe, Ryuhei; Yamaki, Rika.

1989-01-01

Aerosol filters considered so far for nuclear reactor containers in conventional BWR type nuclear power plants make the facility larger and involve a risk of clogging. In view of the above, in the present invention, the diameter of a flow channel of gases entering from a bent pipe to a suppression pool is made smaller thereby decreasing the diameter of gas bubbles in the supperssional pool. Since this reduces the force of surface tension, the diameter of resulted gas bubbles is made remarkably smaller as compared with the case where the gases are released from the lower end of the bent pipe. Since the absorption velocity of bubble-entrained aerosols into water is in proportion to the square of the bubble diameter, the absorption efficiency can be increased remarkably by reducing the diameter of the gas bubbles. Accordingly, it is possible to improve the efficiency of eliminating radioactivity of released gases. (K.M.)

Nuclear reactor building

International Nuclear Information System (INIS)

Oshima, Nobuaki.

1991-01-01

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

Contained fissionly vaporized imploded fission explosive breeder reactor

International Nuclear Information System (INIS)

Marwick, E.F.

1978-01-01

Disclosed is a nuclear reactor system which produces useful thermal power and breeds fissile isotopes wherein large spherical complex slugs containing fissile and fertile isotopes as well as vaporizing and tamping materials are exploded seriatim in a large containing chamber having walls protected from the effects of the explosion by about two thousand tons of slurry of fissile and fertile isotopes in molten alkali metal. The slug which is slightly sub-critical prior to its entry into the centroid portion of the chamber, then becomes slightly more than prompt-critical because of the near proximity of neutron-reflecting atoms and of fissioning atoms within the slurry. The slurry is heated by explosion of the slugs and serves as a working fluid for extraction of heat energy from the reactor. Explosive debris is precipitated from the slurry and used for the fabrication of new slugs

Pressure loadings of Soviet-designed VVER [Water-Cooled, Water-Moderated Energy Reactor] reactor release mitigation structures from large-break LOCAs

International Nuclear Information System (INIS)

Sienicki, J.J.; Horak, W.C.

1989-01-01

Analyses have been carried out of the pressurization of the accident release mitigation structures of Soviet-designed VVER (Water-Cooled, Water-Moderated Energy Reactor) pressurized water reactors following large-break loss-of-coolant accidents. Specific VVER systems for which calculations were performed are the VVER-440 model V230, VVER-440 model V213, and VVER-1000 model V320. Descriptions of the designs of these and other VVER models are contained in the report DOE/NE-0084. The principal objective of the current analyses is to calculate the time dependent pressure loadings inside the accident localization or containment structures immediately following the double-ended guillotine rupture of a primary coolant pipe. In addition, the pressures are compared with the results of calculations of the response of the structures to overpressure. Primary coolant system thermal hydraulic conditions and the fluid conditions at the break location were calculated with the RETRAN-02 Mod2 computer code (Agee, 1984). Pressures and temperatures inside the building accident release mitigation structures were obtained from the PACER (Pressurization Accompanying Coolant Escape from Ruptures) multicompartment containment analysis code developed at Argonne National Laboratory. The analyses were carried out using best estimate models and conditions rather than conservative, bounding-type assumptions. In particular, condensation upon structure and equipment was calculated using correlations based upon analyses of the HDR, Marviken, and Battelle Frankfurt containment loading experiments. The intercompartment flow rates incorporate an effective discharge coefficient and liquid droplet carryover fraction given by expressions of Schwan determined from analyses of the Battelle Frankfurt and Marviken tests. 5 refs., 4 figs

Containment shells of reactor compartments at foreign NPPs

International Nuclear Information System (INIS)

Demidov, A.P.; Savchenko, V.A.

1989-01-01

The modern designes of containment shells (CS) of NPP reactor compartments is described. Much attention is paid to the PCS-3 project envisaging CS inclusion in the complex of NPP passive safety system. The PCS-3 system is developed in the USA for NPP with the improved PWR type reactor. The above system permits to cool the core quickly, to reduce steam pressure in CS down to a safe level and to prevent the discharge of radioactive products in the atmosphere in the case of accidents, even very serious, caused by loss of coolant and core dryout

Pipe line construction for reactor containment buildings

International Nuclear Information System (INIS)

Aoki, Masataka; Yoshinaga, Toshiaki

1978-01-01

Purpose: To prevent the missile phenomenon caused by broken fragments due to pipe whip phenomenon in a portion of pipe lines connected to a reactor containment from prevailing to other portions. Constitution: Various pipe lines connected to the pressure vessel are disposed at the outside of the containments and they are surrounded with a plurality of protection partition walls respectively independent from each other. This can eliminate the effect of missile phenomena upon pipe rupture from prevailing to the pipe lines and instruments. Furthermore this can afford sufficient spaces for the pipe lines, as well as for earthquake-proof supports. (Horiuchi, T.)

Survey of Regulations Applicable to the Finned Containment in Korean Nuclear Power Plant for Light Water Reactor

International Nuclear Information System (INIS)

Noh, Hyung Gyun; Kang, Hie Chan

2016-01-01

In severe accident, the molten corium would discharge into the reactor cavity and interact with water and concrete of cavity. Molten corium includes non-oxidation metals such as Zr, Fe and Cr. These metal species reacted with water emit hydrogen gas. In addition to this, a mount of steam can be emitted to the containment such as steam line break accident. As a result, steam and hydrogen gas can pressurize containment over the design pressure and threaten its integrity. For this reasons, a concept equipped with finned on the containment building was proposed for coping with prolonged accident. Finned containment can enhance heat transfer to the ambient, and the building itself is working as a heat sink. Multiple metal fins and metal rod are penetrated into containment wall, and the rods are working as an additional path of heat removal. To be accepted in the nuclear power plants, this configuration should satisfy the requirement of heat removal and follow all regulations related with containment also. For applying to Korean nuclear power plants, the finned containment should follow all regulations specialized in Korea such as Nuclear regulatory criteria for light water reactor and Guidelines of nuclear safety examination for light water reactor. A concept of containment as a passive cooling system has been proposed. Furthermore, the new containment concept can be applied on the real containment which satisfies the various regulations. Finned containment would be expected positive effects on heat removal from the containment. If the fins are properly welded to the liner, finned containment could satisfy the leak tightness and prevention of external influences. Finned containment could be favorable to protect external impact like aircraft crash because of the additional structural integrity by the fins

Survey of Regulations Applicable to the Finned Containment in Korean Nuclear Power Plant for Light Water Reactor

Energy Technology Data Exchange (ETDEWEB)

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

2016-05-15

In severe accident, the molten corium would discharge into the reactor cavity and interact with water and concrete of cavity. Molten corium includes non-oxidation metals such as Zr, Fe and Cr. These metal species reacted with water emit hydrogen gas. In addition to this, a mount of steam can be emitted to the containment such as steam line break accident. As a result, steam and hydrogen gas can pressurize containment over the design pressure and threaten its integrity. For this reasons, a concept equipped with finned on the containment building was proposed for coping with prolonged accident. Finned containment can enhance heat transfer to the ambient, and the building itself is working as a heat sink. Multiple metal fins and metal rod are penetrated into containment wall, and the rods are working as an additional path of heat removal. To be accepted in the nuclear power plants, this configuration should satisfy the requirement of heat removal and follow all regulations related with containment also. For applying to Korean nuclear power plants, the finned containment should follow all regulations specialized in Korea such as Nuclear regulatory criteria for light water reactor and Guidelines of nuclear safety examination for light water reactor. A concept of containment as a passive cooling system has been proposed. Furthermore, the new containment concept can be applied on the real containment which satisfies the various regulations. Finned containment would be expected positive effects on heat removal from the containment. If the fins are properly welded to the liner, finned containment could satisfy the leak tightness and prevention of external influences. Finned containment could be favorable to protect external impact like aircraft crash because of the additional structural integrity by the fins.

Renewable side reflector structure for a pebble bed high temperature reactor

International Nuclear Information System (INIS)

Martin, Roger.

1977-01-01

The description is given of a renewable side reflector structure for a pebble bed high temperature reactor of the kind comprising a cylindrical graphite vessel constituting the neutron reflector, this vessel being filled with graphite pebbles containing the nuclear fuel and enclosed in a concrete protective containment. The internal peripheral area of the vessel is constituted by a line of adjacent graphite rods mounted so that they can rotate about their longitudinal axis and manoeuvrable from outside the concrete containment by means of a shaft passing into it [fr

Seismic response of the 'Cut-and Cover' type reactor containment considering nonlinear soil behavior

International Nuclear Information System (INIS)

El-Tahan, H.; Reddy, D.V.

1979-01-01

This paper describes some parametric studies of dynamic soil-structure interaction for the 'cut-and-cover' reactor concept. The dynamic loading considered is a horizontal earthquake motion. The high frequency ranges, which must be considered in the study of soil-structure interaction for nuclear power plants, and the nonlinearity of soil behavior during strong earthquakes are adequately taken into account. Soil nonlinearity is accounted for in an approximate manner using a combination of the 'equivalent linear method' and the method of complex response with complex moduli. The structure considered is a reinforced concrete containment for a 1100 - MWe power plant, buried in a dense sand medium. (orig.)

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

Cross-flow electrochemical reactor cells, cross-flow reactors, and use of cross-flow reactors for oxidation reactions

Science.gov (United States)

Balachandran, Uthamalingam; Poeppel, Roger B.; Kleefisch, Mark S.; Kobylinski, Thaddeus P.; Udovich, Carl A.

1994-01-01

This invention discloses cross-flow electrochemical reactor cells containing oxygen permeable materials which have both electron conductivity and oxygen ion conductivity, cross-flow reactors, and electrochemical processes using cross-flow reactor cells having oxygen permeable monolithic cores to control and facilitate transport of oxygen from an oxygen-containing gas stream to oxidation reactions of organic compounds in another gas stream. These cross-flow electrochemical reactors comprise a hollow ceramic blade positioned across a gas stream flow or a stack of crossed hollow ceramic blades containing a channel or channels for flow of gas streams. Each channel has at least one channel wall disposed between a channel and a portion of an outer surface of the ceramic blade, or a common wall with adjacent blades in a stack comprising a gas-impervious mixed metal oxide material of a perovskite structure having electron conductivity and oxygen ion conductivity. The invention includes reactors comprising first and second zones seprated by gas-impervious mixed metal oxide material material having electron conductivity and oxygen ion conductivity. Prefered gas-impervious materials comprise at least one mixed metal oxide having a perovskite structure or perovskite-like structure. The invention includes, also, oxidation processes controlled by using these electrochemical reactors, and these reactions do not require an external source of electrical potential or any external electric circuit for oxidation to proceed.

Ductile fracture behavior of cast structure containing voids

Energy Technology Data Exchange (ETDEWEB)

Gilles, Ph.; Migne, C. [FRAMATOME ANP, 92 - Paris-La-Defence (France); Chapuliot, S. [CEA Saclay, 91 - Gif-sur-Yvette (France). Dept. de Mecanique et de Technologie

2001-07-01

In pressurized water reactors, the primary loop contains cast-piping components made of duplex stainless steel. Due to the presence of ferrite, such steels are susceptible to thermal aging embrittlement, which decrease their fracture resistance. The cast process induces shrinkage cavities, therefore all these components are submitted to liquid penetrant examination and all surface defects are repaired. EDF, CEA and Framatome have conducted experimental and analytical analysis of fatigue and fracture behavior of aged cast stainless steel structures containing shrinkage cavities. The present study considers only ductile tearing and is based on specimen test results and a fracture mechanics model of the interaction between shrinkage cavities. The experimental results presented here show that large groups of shrinkage cavities have almost no influence on the global behavior of the structure. Only for the specimen with the largest reduction of area, a significant reduction of strength has been registered. Using elementary fracture mechanics models, it has been evidenced that failure mechanism of structures containing shrinkage cavities consists in 3 phases: local initiation, macro-crack formation by coalescence and failure by crack instability or collapse depending if J resistance is low or not. No significant changes in global behavior appear in the first phase. (A.C.)

Ductile fracture behavior of cast structure containing voids

International Nuclear Information System (INIS)

Gilles, Ph.; Migne, C.; Chapuliot, S.

2001-01-01

In pressurized water reactors, the primary loop contains cast-piping components made of duplex stainless steel. Due to the presence of ferrite, such steels are susceptible to thermal aging embrittlement, which decrease their fracture resistance. The cast process induces shrinkage cavities, therefore all these components are submitted to liquid penetrant examination and all surface defects are repaired. EDF, CEA and Framatome have conducted experimental and analytical analysis of fatigue and fracture behavior of aged cast stainless steel structures containing shrinkage cavities. The present study considers only ductile tearing and is based on specimen test results and a fracture mechanics model of the interaction between shrinkage cavities. The experimental results presented here show that large groups of shrinkage cavities have almost no influence on the global behavior of the structure. Only for the specimen with the largest reduction of area, a significant reduction of strength has been registered. Using elementary fracture mechanics models, it has been evidenced that failure mechanism of structures containing shrinkage cavities consists in 3 phases: local initiation, macro-crack formation by coalescence and failure by crack instability or collapse depending if J resistance is low or not. No significant changes in global behavior appear in the first phase. (A.C.)

Structural analysis of reactor buildings with help of complete FE models

International Nuclear Information System (INIS)

Diaz, B.E.; Vaz, L.E.; Martha, L.F.R.; Costa, E.

1984-01-01

The reinforced concrete structures located within the steel containment shell of a Reactor Building are formed by highly complex structures subjected to a large amount of actions due to different causes. The analysis of this complex structure can be performed with help of small models, each one representing a part of the global structure. The interaction effects among the partial models are accounted for in approximate way. This approach has been used previously with entire success in the design of 1300 MW PWR nuclear power plants. However a new and entire different approach can be used in the design of these structures. The entire assembly of structural elements of the building is represented and analyzed with help of a single and very large FE model. This paper will present the main characteristics of this type of analysis as well as all the necessary procedures, which must be implemented for the proper data processing of the forces and the automatic reinforced concrete design of the structural elements of the Reactor Building. (Author) [pt

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.

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.)

Three-dimensional fluid-structure interaction dynamics of a pool-reactor in-tank component

International Nuclear Information System (INIS)

Kulak, R.F.

1979-01-01

The safety evaluation of reactor-components often involves the analysis of various types of fluid/structural components interacting in three-dimensional space. For example, in the design of a pool-type reactor several vital in-tank components such as the primary pumps and the intermediate heat exchangers are contained within the primary tank. Typically, these components are suspended from the deck structure and largely submersed in the sodium pool. Because of this positioning these components are vulnerable to structural damage due to pressure wave propagation in the tank during a CDA. In order to assess the structural integrity of these components it is necessary to perform a dynamic analysis in three-dimensional space which accounts for the fluid-structure coupling. A model is developed which has many of the salient features of this fluid-structural component system

Vibration-proof FBR type reactor

International Nuclear Information System (INIS)

Kawamura, Yutaka.

1992-01-01

In a reactor container in an FBR type reactor, an outer building and upper and lower portions of a reactor container are connected by a load transmission device made of a laminated material of rubber and steel plates. Each of the reactor container and the outer building is disposed on a lower raft disposed on a rock by way of a vibration-proof device made of a laminated material of rubber and steel plates. Vibration-proof elements for providing vertical eigen frequency of the vibration-proof system comprising the reactor building and the vibration-proof device within a range of 3Hz to 5Hz are used. That is, the peak of designed acceleration for response spectrum in the horizontal direction of the reactor structural portions is shifted to side of shorter period from the main frequency region of the reactor structure. Alternatively, rigidity of the vibration-proof elements is decreased to shift the peak to the side of long period from the main frequency region. Designed seismic force can be greatly reduced both horizontally and vertically, to reduce the wall thickness of the structural members, improve the plant economy and to ensure the safety against earthquakes. (N.H.)

Verification of Remote Inspection Techniques for Reactor Internal Structures of Liquid Metal Reactor

International Nuclear Information System (INIS)

Joo, Young Sang; Lee, Jae Han

2007-02-01

The reactor internal structures and components of a liquid metal reactor (LMR) are submerged in hot sodium of reactor vessel. The division 3 of ASME code section XI specifies the visual inspection as major in-service inspection (ISI) methods of reactor internal structures and components. Reactor internals of LMR can not be visually examined due to opaque liquid sodium. The under-sodium viewing techniques using an ultrasonic wave should be applied for the visual inspection of reactor internals. Recently, an ultrasonic waveguide sensor with a strip plate has been developed for an application to the under-sodium inspection. In this study, visualization technique, ranging technique and monitoring technique have been suggested for the remote inspection of reactor internals by using the waveguide sensor. The feasibility of these remote inspection techniques using ultrasonic waveguide sensor has been evaluated by an experimental verification

Verification of Remote Inspection Techniques for Reactor Internal Structures of Liquid Metal Reactor

Energy Technology Data Exchange (ETDEWEB)

Joo, Young Sang; Lee, Jae Han

2007-02-15

The reactor internal structures and components of a liquid metal reactor (LMR) are submerged in hot sodium of reactor vessel. The division 3 of ASME code section XI specifies the visual inspection as major in-service inspection (ISI) methods of reactor internal structures and components. Reactor internals of LMR can not be visually examined due to opaque liquid sodium. The under-sodium viewing techniques using an ultrasonic wave should be applied for the visual inspection of reactor internals. Recently, an ultrasonic waveguide sensor with a strip plate has been developed for an application to the under-sodium inspection. In this study, visualization technique, ranging technique and monitoring technique have been suggested for the remote inspection of reactor internals by using the waveguide sensor. The feasibility of these remote inspection techniques using ultrasonic waveguide sensor has been evaluated by an experimental verification.

Material characteristics and construction methods for a typical research reactor concrete containment in Iran

International Nuclear Information System (INIS)

Ebrahimia, Mahsa; Suha, Kune Y.; Eghbalic, Rahman; Jahan, Farzaneh Asadi malek

2012-01-01

Generally selecting an appropriate material and also construction style for a concrete containment due to its function and special geometry play an important role in applicability and also construction cost and duration decrease in a research reactor (RR) project. The reactor containment enclosing the reactor vessel comprises physical barriers reflecting the safety design and construction codes, regulations and standards so as to prevent the community and the environment from uncontrolled release of radioactive materials. It is the third and the last barrier against radioactivity release. It protects the reactor vessel from such external events as earthquake and aircraft crash as well. Thus, it should be designed and constructed in such a manner as to withstand dead and live loads, ground and seismic loads, missiles and aircraft loads, and thermal and shrinkage loads. This study aims to present a construction method for concrete containment of a typical RR in Iran. The work also presents an acceptable characteristic for concrete and reinforcing re bar of a typical concrete containment. The current study has evaluated the various types of the RR containments. The most proper type was selected in accordance with the current knowledge and technology of Iran

Material characteristics and construction methods for a typical research reactor concrete containment in Iran

Energy Technology Data Exchange (ETDEWEB)

Ebrahimia, Mahsa; Suha, Kune Y. [Seoul National Univ., Seoul (Korea, Republic of); Eghbalic, Rahman; Jahan, Farzaneh Asadi malek [School of Architecture and Urbanism, Qazvin (Iran, Islamic Republic of)

2012-10-15

Generally selecting an appropriate material and also construction style for a concrete containment due to its function and special geometry play an important role in applicability and also construction cost and duration decrease in a research reactor (RR) project. The reactor containment enclosing the reactor vessel comprises physical barriers reflecting the safety design and construction codes, regulations and standards so as to prevent the community and the environment from uncontrolled release of radioactive materials. It is the third and the last barrier against radioactivity release. It protects the reactor vessel from such external events as earthquake and aircraft crash as well. Thus, it should be designed and constructed in such a manner as to withstand dead and live loads, ground and seismic loads, missiles and aircraft loads, and thermal and shrinkage loads. This study aims to present a construction method for concrete containment of a typical RR in Iran. The work also presents an acceptable characteristic for concrete and reinforcing re bar of a typical concrete containment. The current study has evaluated the various types of the RR containments. The most proper type was selected in accordance with the current knowledge and technology of Iran.

Nuclear reactor

International Nuclear Information System (INIS)

Hattori, Sadao; Sekine, Katsuhisa.

1987-01-01

Purpose: To decrease the thickness of a reactor container and reduce the height and the height and plate thickness of a roof slab without using mechanical vibration stoppers. Constitution: Earthquake proofness is improved by filling fluids such as liquid metal between a reactor container and a secondary container and connecting the outer surface of the reactor container with the inner surface of the secondary container by means of bellows. That is, for the horizontal seismic vibrations, horizontal loads can be supported by the secondary container without providing mechanical vibration stoppers to the reactor container and the wall thickness can be reduced thereby enabling to simplify thermal insulation structure for the reduction of thermal stresses. Further, for the vertical seismic vibrations, verical loads can be transmitted to the secondary container thereby enabling to reduce the wall thickness in the same manner as for the horizontal load. By the effect of transferring the point of action of the container load applied to the roof slab to the outer circumferential portion, the intended purpose can be attained and, in addition, the radiation dose rate at the upper surface of the roof slab can be decreased. (Kamimura, M.)

Modeling of hydrogen stratification in a pressurized water reactor containment with the contain computer code

International Nuclear Information System (INIS)

Kljenak, I.; Skerlavaj, A.; Parzer, I.

1999-01-01

Hydrogen distribution during a severe accident in a nuclear power plant with a two-loop Westinghouse-type pressurized water reactor was simulated with the CONTAIN computer code. The accidents is initiated by a large-break loss-of-coolant accident which is nit successfully mitigated by the action of the emergency core cooling system. Cases with and without successful actuation of spray systems and fan coolers were considered. The simulations predicted hydrogen stratification within the containment main compartment with intensive hydrogen mixing in the containment dome region. Pressure and temperature responses were analyzed as well.(author)

Structural strength during severe reactor accidents of the VVER- 91 nuclear power plant

International Nuclear Information System (INIS)

Varpasuo, P.

1999-12-01

The report summarises the studies carried out in Fortum Engineering (formerly IVO Power Engineering) between the years of 1992 and 1997 concerning ultimate strength of structures designed to mitigate and contain the consequences of various core melt accident scenarios. The report begins with the description of containment loading situations arising from core melt accidents. These situations are divided to fast and slow loads. Fast loads include ex-vessel steam explosions, steam spikes, hydrogen burns, direct containment heating and missiles. Slow loads are connected with pressure rise inside the containment in case when the containment heat removal system is not functioning. First part of report describes the analyses of reactor cavity based on axi-symmetric load assumptions. These studies are performed with various models like one degree of freedom idealisation, axi-symmetric modelling of geometry and full three-dimensional modelling of geometry. Second part of report describes the analyses of cavity based on non-axi-symmetric load assumptions. Here full 3D- geometry model is used combined with various physical models for the behaviour of reinforced concrete. Third part of report gives short account of the analysis of containment ultimate pressure capacity. The containment model in this case includes pre-stressing tendons and mild steel reinforcing bars. The load is assumed to axi-symmetric internal static pressure. The capacity of the reactor cavity against the ex-vessel steam explosion scenarios for VVER-91 plant concept is established for both axi-symmetric and non-axi-symmetric load models using ANACAP structural analysis code. The validation of the cavity response to ex-vessel steam explosion load using different commercially available codes gave mixed results for both axisymmetric and non-axi-symmetric load presentations.The ultimate static overpressure capacity of the VVER-91 reactor cavity structure was established to be of the order of 10 MPa. This result

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

Swelling of structural materials in fast neutron reactors

International Nuclear Information System (INIS)

Seran, J.L.

1983-06-01

The physical origin of swelling in irradiated materials and the main parameters acting on swelling of SS 316 are examined: temperature, neutron dose, dose rate, chemical composition, strain hardening. Results obtained, in Rapsodie and Phenix reactors, with fuel cans and with the hexagonal tube containing the fuel pins are analyzed and compared with results found in litterature. In conclusion hot swelling of SS 316 is too important at high doses and is will be replaced by austenitic steels stabilized by Ti and ferritic steels or high nickel steels with structural hardening [fr

Development of fast reactor containment safety analysis code, CONTAIN-LMR. (3) Improvement of sodium-concrete reaction model

International Nuclear Information System (INIS)

Kawaguchi, Munemichi; Doi, Daisuke; Seino, Hiroshi; Miyahara, Shinya

2015-01-01

A computer code, CONTAIN-LMR, is an integrated analysis tool to predict the consequence of severe accident in a liquid metal fast reactor. Because a sodium-concrete reaction behavior is one of the most important phenomena in the accident, a Sodium-Limestone Concrete Ablation Model (SLAM) has been developed and installed into the original CONTAIN code at Sandia National Laboratories (SNL) in the U.S. The SLAM treats chemical reaction kinetics between the sodium and the concrete compositions mechanistically using a three-region model, containing a pool (sodium and reaction debris) region, a dry (boundary layer (B/L) and dehydrated concrete) region, and a wet (hydrated concrete) region, the application is limited to the reaction between sodium and limestone concrete. In order to apply SLAM to the reaction between sodium and siliceous concrete which is an ordinary structural concrete in Japan, the chemical reaction kinetics model has been improved to consider the new chemical reactions between sodium and silicon dioxide. The improved model was validated to analyze a series of sodium-concrete experiments which were conducted in Japan Atomic Energy Agency (JAEA). It has been found that relatively good agreement between calculation and experimental results is obtained and the CONTAIN-LMR code has been validated with regard to the sodium-concrete reaction phenomena. (author)

CONTAIN LMR/1B-Mod.1, A computer code for containment analysis of accidents in liquid-metal-cooled nuclear reactors

International Nuclear Information System (INIS)

Murata, K.K.; Carroll, D.E.; Bergeron, K.D.; Valdez, G.D.

1993-01-01

The CONTAIN computer code is a best-estimate, integrated analysis tool for predicting the physical, chemical, and radiological conditions inside a nuclear reactor containment building following the release of core material from the primary system. CONTAIN is supported primarily by the U. S. Nuclear Regulatory Commission (USNRC), and the official code versions produced with this support are intended primarily for the analysis of light water reactors (LWR). The present manual describes CONTAIN LMR/1B-Mod. 1, a code version designed for the analysis of reactors with liquid metal coolant. It is a variant of the official CONTAIN 1.11 LWR code version. Some of the features of CONTAIN-LMR for treating the behavior of liquid metal coolant are in fact present in the LWR code versions but are discussed here rather than in the User's Manual for the LWR versions. These features include models for sodium pool and spray fires. In addition to these models, new or substantially improved models have been installed in CONTAIN-LMR. The latter include models for treating two condensables (sodium and water) simultaneously, sodium atmosphere and pool chemistry, sodium condensation on aerosols, heat transfer from core-debris beds and to sodium pools, and sodium-concrete interactions. A detailed description of each of the above models is given, along with the code input requirements

The status of the Bubbler Condenser Containment System for the Reactors of the VVER-440/213 Type

International Nuclear Information System (INIS)

Karwat, H.; Rosinger, H.E.

1998-01-01

VVER-440/213 Pressurized Water Reactors have a pressure-suppression containment structure called a 'Bubbler Condenser' tower which can reduce the design pressure of the entire containment following a design basis accident (DBA), such as a loss-of-coolant accident (LOCA). The bubbler condenser pressure suppression system provides reduction of the LOCA containment pressure by the condensation of released steam in a water pool. World-wide there are 14 nuclear power plants of the VVER-440/213 type in Eastern Europe and Russia. One of the safety concerns for the VVER-440/213 reactors relates to the ability of the bubbler condenser containment system to function satisfactorily and to maintain its integrity following certain postulated accidents and thus limit the release of radioactive material to the environment. The complicated geometry of the bubbler condenser unit, and the dependence on several moving devices and interlocks are the main doubts expressed by different specialists with regard to the design. General description of the bubbler condenser containment system, the physical processes, concerns and design assessment of the bubbler condenser containment system, presentation of the OECD's Unified Bubbler Condenser Research Project (UBCRP) and the European Commission PHARE/TACIS project. Recent utility investigations are also discussed

TMI in perspective: reactor containment stands up, difficult decisions remain

International Nuclear Information System (INIS)

Corey, G.R.

1979-01-01

Commonwealth Edison Co. is increasing its commitment to nuclear energy after reviewing the performance of the Three Mile Island reactor containment systems. Both the reactor vessel and the secondary containment remained intact and no radiation was reported in the soil or water. The public discussion of energy options which followed the accident will benefit both the public and technical community even if there is a temporary slowdown in nuclear power development. The realities of energy supplies have become evident; i.e., that nuclear and coal are the only available options for the short-term. The discussion should also lead to better personnel training, regulatory reforms, risk-sharing insurance, and international standards. The public hysteria triggered by the accident stemmed partly from the combination of unfortunate incidents and the media coverage, which led to hasty conclusions

ITER containment structures

International Nuclear Information System (INIS)

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

1991-01-01

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

Validation of CONTAIN-LMR code for accident analysis of sodium-cooled fast reactor containments

Energy Technology Data Exchange (ETDEWEB)

Gordeev, S.; Hering, W.; Schikorr, M.; Stieglitz, R. [Inst. for Neutron Physic and Reactor Technology, Karlsruhe Inst. of Technology, Campus Nord (Germany)

2012-07-01

CONTAIN-LMR 1 is an analytical tool for the containment performance of sodium cooled fast reactors. In this code, the modelling for the sodium fire is included: the oxygen diffusion model for the sodium pool fire, and the liquid droplet model for the sodium spray fire. CONTAIN-LMR is also able to model the interaction of liquid sodium with concrete structure. It may be applicable to different concrete compositions. Testing and validation of these models will help to qualify the simulation results. Three experiments with sodium performed in the FAUNA facility at FZK have been used for the validation of CONTAIN-LMR. For pool fire tests, calculations have been performed with two models. The first model consists of one gas cell representing the volume of the burn compartment. The volume of the second model is subdivided into 32 coupled gas cells. The agreement between calculations and experimental data is acceptable. The detailed pool fire model shows less deviation from experiments. In the spray fire, the direct heating from the sodium burning in the media is dominant. Therefore, single cell modeling is enough to describe the phenomena. Calculation results have reasonable agreement with experimental data. Limitations of the implemented spray model can cause the overestimation of predicted pressure and temperature in the cell atmosphere. The ability of the CONTAIN-LMR to simulate the sodium pool fire accompanied by sodium-concrete reactions was tested using the experimental study of sodium-concrete interactions for construction concrete as well as for shielding concrete. The model provides a reasonably good representation of chemical processes during sodium-concrete interaction. The comparison of time-temperature profiles of sodium and concrete shows, that the model requires modifications for predictions of the test results. (authors)

Positron annihilation studies on structural materials for nuclear reactors

International Nuclear Information System (INIS)

Rajaraman, R.; Amarendra, G.; Sundar, C.S.

2012-01-01

Structural steels for nuclear reactors have renewed interest owing to the future advanced fission reactor design with increased burn-up goals as well as for fusion reactor applications. While modified austenitic steels continue to be the main cladding materials for fast breeder reactors, Ferritic/martensitic steels and oxide dispersion strengthened ferritic steels are the candidate materials for future reactors applications in India. Sensitivity and selectivity of positron annihilation spectroscopy to open volume type defects and nano clusters have been extensively utilized in studying reactor materials. We have recently reviewed the application of positron techniques to reactor structural steels. In this talk, we will present successful application of positron annihilation spectroscopy to probe various structural materials such as D9, ferritic/martensitic, oxide dispersion strengthened (ODS) steels and related model alloys, highlighting our recent studies. (author)

Plant life management of the ACR-1000 Concrete containment structure

International Nuclear Information System (INIS)

Abrishami, H.H.; Ricciuti, R.; Elgohary, M.

2009-01-01

The Ageing of reinforced concrete structures due to service conditions, aggressive environments, or accidents may cause their strength, serviceability and durability to decrease over time. For a new plant, a Plant Life Management (PLiM) program should start in the design process and then continues through the plant operation and decommissioning. Hence, PLiM must provide not only Ageing Management program (AMP) but also provide requirements on material characteristic and design criteria as well. The purpose of this paper is to present the Plant Life Management (PLiM) strategy for the concrete containment structure of the ACR-10001 (Advanced CANDU Reactor) designed by AECL. The ACR-1000 is designed for a 100-year plant life including 60-year operating life and an additional 40-year decommissioning period. The approach adopted for the PLiM strategy of the concrete containment structure is a preventive one, key areas being: 1) design methodology, 2) material performance and 3) ageing management program. During the design phase, in addition to strength and serviceability, durability, throughout the service life and decommissioning phase of the ACR-1000 structure, is a major consideration. Factors affecting durability design include: a) concrete performance, b) structural application, and c) consideration of environmental conditions. In addition to addressing the design methodology and material performance requirements, a systematic approach for the ageing management program for the concrete containment structure is presented. (authors)

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

Preliminary conceptual design and analysis on KALIMER reactor structures

International Nuclear Information System (INIS)

Kim, Jong Bum

1996-10-01

The objectives of this study are to perform preliminary conceptual design and structural analyses for KALIMER (Korea Advanced Liquid Metal Reactor) reactor structures to assess the design feasibility and to identify detailed analysis requirements. KALIMER thermal hydraulic system analysis results and neutronic analysis results are not available at present, only-limited preliminary structural analyses have been performed with the assumptions on the thermal loads. The responses of reactor vessel and reactor internal structures were based on the temperature difference of core inlet and outlet and on engineering judgments. Thermal stresses from the assumed temperatures were calculated using ANSYS code through parametric finite element heat transfer and elastic stress analyses. While, based on the results of preliminary conceptual design and structural analyses, the ASME Code limits for the reactor structures were satisfied for the pressure boundary, the needs for inelastic analyses were indicated for evaluation of design adequacy of the support barrel and the thermal liner. To reduce thermal striping effects in the bottom are of UIS due to up-flowing sodium form reactor core, installation of Inconel-718 liner to the bottom area was proposed, and to mitigate thermal shock loads, additional stainless steel liner was also suggested. The design feasibilities of these were validated through simplified preliminary analyses. In conceptual design phase, the implementation of these results will be made for the design of the reactor structures and the reactor internal structures in conjunction with the thermal hydraulic, neutronic, and seismic analyses results. 4 tabs., 24 figs., 4 refs. (Author)

A remote inspection system for use inside reactor containment vessels

International Nuclear Information System (INIS)

Aoki, Toshihiko; Kashiwai, Jun-ichi; Yamamoto, Ikuo; Fukada, Koichi; Yamanaka, Yoshinobu.

1985-01-01

The harsh environment in the reactor-containment vesels of pressurized-water reactor nuclear-power plants precludes the possibility of direct circuit inspection; a remote-inspection system is essential. A robot for performing this task must not only be able to withstand the harsh conditions but must also be small and maneuverable enough to function effectively within complex and confined spaces. The article describes a monorail-type remote-inspection robot developed by Mitsubishi Electric to meet these needs, which is now under trial production and testing. (author)

Concrete containment tests: Phase 2, Structural elements with liner plates: Interim report

International Nuclear Information System (INIS)

Hanson, N.W.; Roller, J.J.; Schultz, D.M.; Julien, J.T.; Weinmann, T.L.

1987-08-01

The tests described in this report are part of Phase 2 of the Electric Power Research Institute (EPRI) program. The overall objective of the EPRI program is to provide a test-verified analytical method of estimating capacities of concrete reactor containment buildings under internal overpressurization from postulated degraded core accidents. The Phase 2 testing included seven large-scale specimens representing structural elements from reinforced and prestressed concrete reactor containment buildings. Six of the seven test specimens were square wall elements. Of these six specimens, four were used for biaxial tension tests to determine strength, deformation, and leak-rate characteristics of full-scale wall elements representing prestressed concrete containment design. The remaining two square wall elements were used for thermal buckling tests to determine whether buckling of the steel liner plate would occur between anchorages when subjected to a sudden extreme temperature differential. The last of the seven test specimens for Phase 2 represented the region where the wall and the basemat intersect in a prestressed concrete containment building. A multi-directional loading scheme was used to produce high bending moments and shear in the wall/basemat junction region. The objective of this test was to determine if there is potential for liner plate tearing in the junction region. Results presented include observed behavior and extensive measurements of deformations and strains as a function of applied load. The data are being used to confirm analytical models for predicting strength and deformation of containment structures in a separate parallel analytical investigation sponsored by EPRI

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

Optically monitoring device in reactor container

International Nuclear Information System (INIS)

Takeuchi, Tsutomu; Kawamoto, Kikuo.

1993-01-01

In the device of the present invention, cable penetrations are necessary for transmission from a great number of electrical instrumentations disposed in a reactor container to the outside. Optical cables are passed through the cable penetrations to use optical signals for signal transmission. That is, pulses are injected from one end of the optical cable and a specific light (raman scattered light) among reflected lights, after elapse of a predetermined time, is measured. With such procedures, temperature at the reflection point can be measured. In this case, if emission and discrimination of the pulses are conducted in a time sharing fashion, the temperature is measured as an average value for the 1m length corresponding to the time determined by the limit. Accordingly, greater number of temperature measuring points than that in the prior art (there is a reactor which has about 170 points) are measured by a lesser number of cables (one at minimum). For instruments used for other than temperature, if the device is diagnosed by itself, by making the constitution intelligent and utilizing optical output, reliability for the measurement can be improved. (I.S.)

Axisymmetric global structural analysis of BARC prestressed concrete containment model for beyond design pressure

International Nuclear Information System (INIS)

Singh, Tarvinder; Singh, R.K.; Ghosh, A.K.

2008-10-01

In order to check the adequacy of the Indian Pressurized Heavy Water Reactor (PHWR) containment structure to withstand severe accident induced internal pressure load, the ultimate load capacity assessment is required. Reactor Safety Division (RSD) of Bhabha Atomic Research Centre (BARC) has initiated an experimental program at BARC Tarapur Containment Test Facility to evaluate the ultimate load capacity of Indian PHWR containment. For this study, BARC Containment Model (BARCOM), which is 1:4 scale representation of Tarapur Atomic Power Station (TAPS) unit-3 and 4 540 MWe PHWR Inner Containment of Pre-stressed Concrete has been constructed. The model includes all the important major design features of the prototype containment and simulates Main Air Lock (MAL), Steam Generator (SG), Emergency Air Lock (EAL) and Fueling Machine Air Lock (FMAL) openings. The design pressure (Pd) of BARCOM is 1.44kg/cm 2 (g), which is same as the prototype. The pretest analysis of BARCOM has been performed with finite element axi-symmetric modeling. The objective of this simulation was to understand the behavior of containment model under internal pressure and find out the various failure modes and critical locations important for instrumentation during the experiment. The structural response of the containment model is assessed in terms of wall and dome displacement; cracking of concrete, longitudinal and hoop strains and stresses. Another objective of the analysis was to predict the various failure modes of BARCOM with regard to the concrete cracking, reinforcement yielding and tendon inelastic behavior along with the estimation of the ultimate load capacity of the containment model. It is noted that the BARCOM has an ultimate load capacity factor of 3.54 Pd. However, further analysis is needed to quantify the factor of safety with detail 3D model, which should account for the local structural behavior due to various openings. Meanwhile, this preliminary simplified analysis helps to

Damage evaluation of 500 MWe Indian Pressurized Heavy Water Reactor nuclear containment for aircraft impact

Energy Technology Data Exchange (ETDEWEB)

Kukreja, Mukesh [Reactor Safety Division, Bhabha Atomic Research Centre, Trombay, Mumbai 400085 (India)]. E-mail: mrkukreja@yahoo.com

2005-08-01

Safety assessment of Indian nuclear containments has been carried out for aircraft impact. The loading time history for Boeing and Airbus categories of aircrafts is generated based on the principle of momentum transfer of crushable aircrafts. The case studies include the analysis of BWR Mark III containment as a benchmark problem and analyses of Pressurised Heavy Water Reactor containment (inner and outer containment) for impulsive loading due to aircraft impact. Initially, the load is applied on outer containment wall model and subsequently the load is transferred to inner containment after the local perforation of the outer containment wall is noticed in the transient simulation. The analysis methodology evolved in the present work would be useful for studying the behavior of double containment walls and multi barrier structural configurations for aircraft impact with higher energies. The present analysis illustrates that with the provision of double containments for Indian nuclear power plants, adequate reserve strength is available for the case of an extremely low probability event of missile impact generated due to the commercial aircrafts operated in India.

Damage evaluation of 500 MWe Indian Pressurized Heavy Water Reactor nuclear containment for aircraft impact

International Nuclear Information System (INIS)

Kukreja, Mukesh

2005-01-01

Safety assessment of Indian nuclear containments has been carried out for aircraft impact. The loading time history for Boeing and Airbus categories of aircrafts is generated based on the principle of momentum transfer of crushable aircrafts. The case studies include the analysis of BWR Mark III containment as a benchmark problem and analyses of Pressurised Heavy Water Reactor containment (inner and outer containment) for impulsive loading due to aircraft impact. Initially, the load is applied on outer containment wall model and subsequently the load is transferred to inner containment after the local perforation of the outer containment wall is noticed in the transient simulation. The analysis methodology evolved in the present work would be useful for studying the behavior of double containment walls and multi barrier structural configurations for aircraft impact with higher energies. The present analysis illustrates that with the provision of double containments for Indian nuclear power plants, adequate reserve strength is available for the case of an extremely low probability event of missile impact generated due to the commercial aircrafts operated in India

Studies for reactor containment at Narora

International Nuclear Information System (INIS)

Kurien, M.Z.

1975-01-01

The primary reactor containment at Narora has the unusual function of also carrying the heavy boilers (heat exchangers) and providing a flat work surface over the roof. The site being in a seismically active region, certain conflicting requirements in regard to ideal conditions to resist and analyse functional loads and seismic loads have to be reconciled. Alternative schemes are proposed and the stress flow for each major loading case examined and their consequent relative advantages discussed. Problems at the joints and functions are examined and the implications of various arrangements are reviewed. Suggestions to reduce the magnitude of the problems are tentatively proposed. (author)

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

International Nuclear Information System (INIS)

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

1978-01-01

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

Reactor building with internal structure of which the movements are independent of those of the general raft and process for building these internal structures

International Nuclear Information System (INIS)

Hista, J.C.

1982-01-01

This reactor building includes a containment enclosure for the internal structures composed of a slab wedged on its periphery against the containment enclosure gusset and resting on the general raft by means of a peripheral bearing ring, a compressible layer being provided between the general raft and the slab [fr

Core catcher for nuclear reactor core meltdown containment

International Nuclear Information System (INIS)

Driscoll, M.J.; Bowman, F.L.

1978-01-01

A bed of graphite particles is placed beneath a nuclear reactor core outside the pressure vessel but within the containment building to catch the core debris in the event of failure of the emergency core cooling system. Spray cooling of the debris and graphite particles together with draining and flooding of coolant fluid of the graphite bed is provided to prevent debris slump-through to the bottom of the bed

REACTOR FUEL ELEMENTS TESTING CONTAINER

Science.gov (United States)

Whitham, G.K.; Smith, R.R.

1963-01-15

This patent shows a method for detecting leaks in jacketed fuel elements. The element is placed in a sealed tank within a nuclear reactor, and, while the reactor operates, the element is sparged with gas. The gas is then led outside the reactor and monitored for radioactive Xe or Kr. (AEC)

Management of research reactor; dynamic characteristics analysis for reactor structures related with vibration of HANARO fuel assembly

Energy Technology Data Exchange (ETDEWEB)

Ahn, Chang Kee; Shim, Joo Sup [Shinwa Technology Information, Seoul (Korea)

2001-04-01

The objective of this study is to deduce the dynamic correlation between the fuel assembly and the reactor structure. Dynamic characteristics analyses for reactor structure related with vibration of HANARO fuel assembly have been performed For the dynamic characteristic analysis, the in-air models of the round and hexagonal flow tubes, 18-element and 36-element fuel assemblies, and reactor structure were developed. By calculating the hydrodynamic mass and distributing it on the in-air models, the in-water models of the flow tubes, the fuel assemblies, and the reactor structure were developed. Then, modal analyses for developed in-air and in-water models have been performed. Especially, two 18-element fuel assemblies and three 36-element fuel assemblies were included in the in-water reactor models. For the verification of the modal analysis results, the natural frequencies and the mode shapes of the fuel assembly were compared with those obtained from the experiment. Finally the analysis results of the reactor structure were compared with them performed by AECL Based on the reactor model without PCS piping, the in-water reactor model including the fuel assemblies was developed, and its modal analysis was performed. The analysis results demonstrate that there are no resonance between the fuel assembly and the reactor structures. 26 refs., 419 figs., 85 tabs. (Author)

Assessment of structural materials inside the reactor pool of the Dalat research reactor

International Nuclear Information System (INIS)

Nguyen Nhi Dien; Luong Ba Vien; Nguyen Minh Tuan; Trang Cao Su

2010-01-01

Originally the Dalat Nuclear Research Reactor (DNRR) was a 250-kW TRIGA MARK II reactor, started building from early 1960s and achieved the first criticality on February 26, 1963. During the 1982-1984 period, the reactor was reconstructed and upgraded to 500kW, and restarted operation on March 20, 1984. From the original TRIGA reactor, only the pool liner, beam ports, thermal columns, and graphite reflector have been remained. The structural materials of pool liner and other components of TRIGA were made of aluminum alloy 6061 and aluminum cladding fuel assemblies. Some other parts, such as reactor core, irradiation rotary rack around the core, vertical irradiation facilities, etc. were replaced by the former Soviet Union's design with structural materials of aluminum alloy CAV-1. WWR-M2 fuel assemblies of U-Al alloy 36% and 19.75% 235 U enrichment and aluminum cladding have been used. In its original version, the reactor was setting upon an all-welded aluminum frame supported by four legs attached to the bottom of the pool. After the modification made, the new core is now suspended from the top of the pool liner by means of three aluminum concentric cylindrical shells. The upper one has a diameter of 1.9m, a length of 3.5m and a thickness of 10mm. This shell prevents from any visual access to the upper part of the pool liner, but is provided with some holes to facilitate water circulation in the 4cm gap between itself and the reactor pool liner. The lower cylindrical shells act as an extracting well for water circulation. As reactor has been operated at low power of 500 kW, it was no any problem with degradation of core structural materials due to neutron irradiation and thermal heat, but there are some ageing issues with aluminum liner and other structures (for example, corrosion of tightening-up steel bolt in the fourth beam port and flood of neutron detector housing) inside the reactor pool. In this report, the authors give an overview and assessment of

Structural design of a shipping container of fuel elements, non-irradiated, for research reactors

International Nuclear Information System (INIS)

Morales Uzqueda, Eduardo Mario

2013-01-01

This work is part of a project whose ultimate goal the creation and subsequent discharge of a transport container fuel assemblies for use by the Chilean Commission for Energy Nuclear. In principle it is covered in the design stage, considering the materials and methods used, to further develop a stage of checking voltages in the container to be manufactured. To achieve the first phase of the study is necessary to understand and warn the importance, geometry and content of the fuel elements to be transported, for which there are standards that provide fundamental material for proper classification of both content and container design. Once approved the design of the structure is critical examine both in normal operation and in the case of accidents that are established by international bodies. for appropriate analytical methods that seek to achieve is use a appropriate representation of the behavior of the structure. in addition to strengthen the theory computer simulations of the tests used applied, where the results will be contrasted with the first method of calculation. Results are obtained for the stress field and displacement total delivering the information necessary to approve the container

State of the art seismic analysis for CANDU reactor structure components using condensation method

Energy Technology Data Exchange (ETDEWEB)

Soliman, S A; Ibraham, A M; Hodgson, S [Atomic Energy of Canada Ltd., Saskatoon, SK (Canada)

1996-12-31

The reactor structure assembly seismic analysis is a relatively complex process because of the intricate geometry with many different discontinuities, and due to the hydraulic attached mass which follows the structure during its vibration. In order to simulate reasonably accurate behaviour of the reactor structure assembly, detailed finite element models are generated and used for both modal and stress analysis. Guyan reduction condensation method was used in the analysis. The attached mass, which includes the fluid mass contained in the components plus the added mass which accounts for the inertia of the surrounding fluid entrained by the accelerating structure immersed in the fluid, was calculated and attached to the vibrating structures. The masses of the attached components, supported partly or totally by the assembly which includes piping, reactivity control units, end fittings, etc. are also considered in the analysis. (author). 4 refs., 6 tabs., 4 figs.

Estimation of radioactivity in structural materials of ETRR-1 reactor

Energy Technology Data Exchange (ETDEWEB)

Imam, M [National Center for Nuclear Safety and Radiation Control Atomic Energy Authority, Cairo (Egypt)

1997-12-31

Precise knowledge of the thermal neutron flux in the different structural materials of a reactor is necessary to estimate the radioactive inventory in these materials that are needed in any decommissioning study of the reactor. ETRR-1 is a research reactor that went critical on 2/1691. In spite of this long age of the reactor, the effective operation time of this reactor is very short since the reactor was shutdown for long periods. Because of this long age one may think of reactor decommissioning. For this purpose, the radioactivity of the reactor structural materials was estimated. Apart from the reactor core, the important structural materials in the ETRR-1 are the reactor tank, shielding concrete, and the graphite thermal column. The thermal neutron flux was determined by the monte Carlo method in these materials and the isotope inventory and the radioactivity were calculated by the international code ORIGEN-JR. 1 fig.

Nuclear reactor fuel structure containing uranium alloy wires embedded in a metallic matrix plate

Science.gov (United States)

Travelli, Armando

1988-01-01

A flat or curved plate structure, to be used as fuel in a nuclear reactor, comprises elongated fissionable wires or strips embedded in a metallic continuous non-fissionable matrix plate. The wires or strips are made predominantly of a malleable uranium alloy, such as uranium silicide, uranium gallide or uranium germanide. The matrix plate is made predominantly of aluminum or an aluminum alloy. The wires or strips are located in a single row at the midsurface of the plate, parallel with one another and with the length dimension of the plate. The wires or strips are separated from each other, and from the surface of the plate, by sufficient thicknesses of matrix material, to provide structural integrity and effective fission product retention, under neutron irradiation. This construction makes it safely feasible to provide a high uranium density, so that the uranium enrichment with uranium 235 may be reduced below about 20%, to deter the reprocessing of the uranium for use in nuclear weapons.

Characterizing the biotransformation of sulfur-containing wastes in simulated landfill reactors.

Science.gov (United States)

Sun, Wenjie; Sun, Mei; Barlaz, Morton A

2016-07-01

Landfills that accept municipal solid waste (MSW) in the U.S. may also accept a number of sulfur-containing wastes including residues from coal or MSW combustion, and construction and demolition (C&D) waste. Under anaerobic conditions that dominate landfills, microbially mediated processes can convert sulfate to hydrogen sulfide (H2S). The presence of H2S in landfill gas is problematic for several reasons including its low odor threshold, human toxicity, and corrosive nature. The objective of this study was to develop and demonstrate a laboratory-scale reactor method to measure the H2S production potential of a range of sulfur-containing wastes. The H2S production potential was measured in 8-L reactors that were filled with a mixture of the target waste, newsprint as a source of organic carbon required for microbial sulfate reduction, and leachate from decomposed residential MSW as an inoculum. Reactors were operated with and without N2 sparging through the reactors, which was designed to reduce H2S accumulation and toxicity. Both H2S and CH4 yields were consistently higher in reactors that were sparged with N2 although the magnitude of the effect varied. The laboratory-measured first order decay rate constants for H2S and CH4 production were used to estimate constants that were applicable in landfills. The estimated constants ranged from 0.11yr(-1) for C&D fines to 0.38yr(-1) for a mixed fly ash and bottom ash from MSW combustion. Copyright © 2016 Elsevier Ltd. All rights reserved.

Deposition of RuO 4 on various surfaces in a nuclear reactor containment

Science.gov (United States)

Holm, Joachim; Glänneskog, Henrik; Ekberg, Christian

2009-07-01

During a severe nuclear reactor accident with air ingress, ruthenium can be released from the nuclear fuel in the form of ruthenium tetroxide. Hence, it is important to investigate how the reactor containment is able to reduce the source term of ruthenium. The aim of this work was to investigate the deposition of gaseous ruthenium tetroxide on aluminium, copper and zinc, which all appear in relatively large amounts in reactor containment. The experiments show that ruthenium tetroxide is deposited on all the metal surfaces, especially on the copper and zinc surfaces. A large deposition of ruthenium tetroxide also appeared on the relatively inert glass surfaces in the experimental set-ups. The analyses of the different surfaces, with several analytical methods, showed that the form of deposited ruthenium was mainly ruthenium dioxide.

Passive containment cooling system performance in the simplified boiling water reactor

International Nuclear Information System (INIS)

Shiralkar, B.S.; Gamble, R.E.; Yadigaroglu, G.

1997-01-01

The Simplified Boiling Water Reactor (SBWR) incorporates a passive system for decay heat removal from the containment in the event of a postulated Loss-of-Coolant Accident (LOCA). Decay heat is removed by condensation of the steam discharged from the reactor pressure vessel (RPV) in three condensers which comprise the Passive Containment Cooling System (PCCS). These condensers are designed to carry the heat load while transporting a mixture of steam and noncondensible gas (primarily nitrogen) from the drywell to the suppression chamber. This paper describes the expected LOCA response of the SBWR with respect to the PCCS performance, based on analysis and test results. The results confirm that the PCCS has excess capacity for decay heat removal and that overall system performance is very robust. 12 refs., 8 figs

Nuclear reactor constructions

International Nuclear Information System (INIS)

Aspden, G.J.

1980-01-01

A nuclear reactor construction comprising a reactor core submerged in a pool of liquid metal coolant in a primary vessel which is suspended from the roof structure of a containment vault. Control rods supported from the roof structure are insertable in the core which is carried on a support structure from the wall of the primary vessel. To prevent excessive relaxation of the support structure whereby the control rods would be displaced relative to the core, the support structure incorporates a normally inactive secondary structure designed to become effective in bracing the primary structure against further relaxation beyond a predetermined limit. (author)

Industrial structure at research reactor suppliers

International Nuclear Information System (INIS)

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

2001-01-01

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

Prestressed and reinforced concrete containments. Analysis - design - construction

International Nuclear Information System (INIS)

Schnellenbach, G.

1975-01-01

Nuclear reactors performing in the German Federal Republic to date were supplied with steel containments. The first reinforced concrete and prestressed concrete containments, respectively, are going to be used for the nuclear power plants Kalkar and Gundremmingen (KRB II) as well as for the HTR plant. Because of their function and nature of loading these structures, similarly to the prestressed concrete reactor pressure vessels, belong to the special structures of civil engineering. Yet, they are substantially different from the prestressed concrete reactor pressure vessels. The problems connected with analysis, design, and construction of these structures are new as well. (orig.) [de

Salient design features of secondary containment structure of Narora Atomic Power Project

International Nuclear Information System (INIS)

Rahalkar, B.D.

1975-01-01

Design of the secondary containment structure for Narora Atomic Power Project is an improvement over the two earlier structures at of Rajasthan and Kalpakkam wherein Candu-type of reactors are involved. The major improvements envisaged are : to limit the leakage through the double containment envelope to 0.1% of volume of the building per day as against 0.1% per hour achieved for earlier stations; to separate heavy water atmosphere from that of light water for effective heavy water recovery; and better man-rem budgetting by limiting inner containment structure upto boiler room floor level and making boiler room area accessible during normal operation for servicing of light water system equipment. Narora Atomic Power Station is located in the Indo-Gangetic alluvial plains in seismically active zone IV. Comprehensive soil investigation, including dynamic properties of soil is required to be undertaken as the foundation level of the containment structure is 17 M below the ground level. The salient results of this investigation relevant to the foundations as well as type of foundation proposed are presented in brief. Double containment concept similar to that adopted for Kalpakkam station is provided for this station also. However, necessary changes in design to withstand large earthquake forces are required to be made. These design problems are discussed in brief. (author)

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.)

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

Science.gov (United States)

Hill, P.R.

1994-12-27

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

Aging of concrete containment structures in nuclear power plants

International Nuclear Information System (INIS)

Naus, D.J.; Oland, C.B.; Ellingwood, B.; Mori, Yasuhiro; Arndt, E.G.

1992-01-01

Concrete structures play a vital role in the safe operation of all light-water reactor plants in the US Pertinent concrete structures are described in terms of their importance design, considerations, and materials of construction. Degradation factors which can potentially impact the ability of these structures to meet their functional and performance requirements are identified. Current inservice inspection requirements for concrete containments are summarized. A review of the performance history of the concrete components in nuclear power plants is provided. A summary is presented. A summary is presented of the Structural Aging (SAG) Program being conducted at the Oak Ridge National Laboratory for the US Nuclear Regulatory Commission. The SAG Program is addressing the aging management of safety-related concrete structures in nuclear power plants for the purpose of providing improved bases for their continued service. The program consists of a management task and three technical tasks: materials property data base, structural component assessment/repair technologies, and quantitiative methodology for continued service conditions. Objectives and a summary of accomplishments under each of these tasks are presented

Structural mechanics and reactor safety

International Nuclear Information System (INIS)

Brandes, K.

1983-01-01

Operational safety and reliability of nuclear power plants widely depend on the mechanical behaviour of their structural components and their resistance to the various and complex influences. Durability and consistency of structural components are determined by the kind of strain - during the life - and by environmental conditions. The Conferences on Structural Mechanics in Reactor Technology (SMiRT) are dedicated to the discussion of such questions. The 7th of these Conferences taking place in 2-year increments was held in Chicago in August 1983. The number of contributions again increased, the number of participants slightly decreased. There are some trends in this field worth mentioning, in particular the fact that experience from design and operation of nuclear power plants now available is more and more made use of, and that more and more attention is given the problems of fusion reactors. (orig./HP) [de

The containment safety of the Dragon Reactor

International Nuclear Information System (INIS)

Cullington, G.R.

1967-08-01

The original design of the Dragon Reactor was based upon the assumption that fission product emitting fuel elements would be used, leading to two significant considerations. First, a highly active primary circuit would result in normal operation, and second, under accident conditions involving massive core damage and corrosion following a major pressure vessel failure, the bulk of the core burden of fission products would be released. The adoption of coated particle fuel able to retain fission products has changed significantly the philosophy behind the design of the containment. The new philosophy is described and its effect on operating principles is discussed. (UK)

A proposed structural, risk-informed approach to the periodicity of CANDU-6 nuclear containment integrated leak rate testing

Energy Technology Data Exchange (ETDEWEB)

Saliba, N. [McGill Univ., Dept. of Civil Engineering and Applied Mechanics, Montreal, Quebec (Canada); Komljenovic, D. [Hydro-Quebec, Gentilly-2 Nuclear Power Plant, Becancour, Quebec (Canada); Chouinard, L. [McGill Univ., Dept. of Civil Engineering and Applied Mechanics, Montreal, Quebec (Canada); Vaillancourt, R.; Chretien, G. [Hydro-Quebec, Gentilly-2 Nuclear Power Plant, Becancour, Quebec (Canada); Gocevski, V. [Hydro-Quebec Equipements, Montreal, Quebec (Canada)

2010-07-01

As ultimate lines of defense against leakage of large amounts of radioactive material to the environment in case of major reactor accidents, containments have been monitored through well designed periodic tests to ensure their proper performance. Regulatory organizations have imposed types and frequencies of containment tests based on highly-conservative deterministic approaches, and judgments of knowledgeable experts. Recent developments in the perception and methods of risk evaluation have been applied to rationalize the leakage-rate testing frequencies while maintaining risks within acceptable levels, preserving the integrity of containments, and respecting the defense-in-depth philosophy. The objective of this paper is to introduce a proposed risk-informed decision making framework on the periodicity of nuclear containment ILRTs for CANDU-6 nuclear power plants based on five main decision criteria, namely: 1) the containment structural integrity; 2) inputs from PSA Level-2; 3) the requirements of deterministic safety analyses and defense-in-depth concepts; 4- the obligations under regulatory and standard requirements; and 5) the return of experience from nuclear containments historic performance. The concepts of dormant reliability and structural fragility will guide the assessment of the containment structural integrity, within the general context of a global containment life cycle management program. This study is oriented towards the requirements of CANDU-6 reactors, in general, and Hydro-Quebec's Gentilly-2 nuclear power plant, in particular. The present article is the first part in a series of papers that will comprehensively detail the proposed research. (author)

A proposed structural, risk-informed approach to the periodicity of CANDU-6 nuclear containment integrated leak rate testing

International Nuclear Information System (INIS)

Saliba, N.; Komljenovic, D.; Chouinard, L.; Vaillancourt, R.; Chretien, G.; Gocevski, V.

2010-01-01

As ultimate lines of defense against leakage of large amounts of radioactive material to the environment in case of major reactor accidents, containments have been monitored through well designed periodic tests to ensure their proper performance. Regulatory organizations have imposed types and frequencies of containment tests based on highly-conservative deterministic approaches, and judgments of knowledgeable experts. Recent developments in the perception and methods of risk evaluation have been applied to rationalize the leakage-rate testing frequencies while maintaining risks within acceptable levels, preserving the integrity of containments, and respecting the defense-in-depth philosophy. The objective of this paper is to introduce a proposed risk-informed decision making framework on the periodicity of nuclear containment ILRTs for CANDU-6 nuclear power plants based on five main decision criteria, namely: 1) the containment structural integrity; 2) inputs from PSA Level-2; 3) the requirements of deterministic safety analyses and defense-in-depth concepts; 4- the obligations under regulatory and standard requirements; and 5) the return of experience from nuclear containments historic performance. The concepts of dormant reliability and structural fragility will guide the assessment of the containment structural integrity, within the general context of a global containment life cycle management program. This study is oriented towards the requirements of CANDU-6 reactors, in general, and Hydro-Quebec's Gentilly-2 nuclear power plant, in particular. The present article is the first part in a series of papers that will comprehensively detail the proposed research. (author)

Reactor container

International Nuclear Information System (INIS)

Oikawa, Hirohide; Otonari, Jun-ichiro; Tozaki, Yuka.

1993-01-01

Partition walls are disposed between a reactor pressure vessel and a suppression chamber to separate a dry well to an upper portion and a lower portion. A communication pipe is disposed to the partition walls. One end of the communication pipe is opened in an upper portion of the dry well at a position higher than a hole disposed to a bent tube of the suppression chamber. When coolants overflow from a depressurization valve by an erroneous operation of an emergency reactor core cooling device, the coolants accumulate in the upper portion of the dry well. When the pipeline is ruptured at the upper portion of the pressure vessel, only the inside of the pressure vessel and the upper portion of the dry well are submerged in water. In this case, the water level of the coolants does not elevate to the opening of the commuication pipe but they flow into the suppression chamber from the hole disposed to the bent tube. Since the coolants do not flow out to the lower portion of the dry well, important equipments such as control rod drives disposed at the lower portion of the dry wall can be prevented from submerging in water. (I.N.)

Effects of aging in containment spray injection system of PWR reactor containment

International Nuclear Information System (INIS)

Borges, Diogo da S.; Lava, Deise D.; Affonso, Renato R.W.; Guimaraes, Antonio C.F.; Moreira, Maria de L.

2014-01-01

This paper presents a contribution to the study of the components aging process in commercial plants of Pressurized Water Reactors (PWR). The analysis is done by applying the method of Fault trees, Monte Carlo Method and Fussell-Vesely Importance Measurement. The study on the aging of nuclear plants, is related to economic factors involved directly with the extent of their operational life, and also provides important data on issues of safety. The most recent case involving the process of extending the life of a PWR plant can be seen in Angra 1 Nuclear Power Plant by investing $ 27 million in the installation of a new reactor cover. The corrective action generated an extension of the useful life of Angra 1 estimated in twenty years, and a great savings compared to the cost of building a new plant and the decommissioning of the first, if it had reached the operation time out 40 years. The extension of the lifetime of a nuclear power plant must be accompanied by special attention from the most sensitive components of the systems to the aging process. After the application of the methodology (aging analysis of Containment Spray Injection System (CSIS)) proposed in this paper, it can be seen that increasing the probability of failure of each component, due to the aging process, generate an increased general unavailability of the system that contains these basic components. The final results obtained were as expected and can contribute to the maintenance policy, preventing premature aging in nuclear power systems

Fuel transporting device in nuclear reactor

International Nuclear Information System (INIS)

Inoue, Tatsumi.

1975-01-01

Object: To obtain a support structure of an excellent quakeproof property for a fuel transporting device provided for the transportation of fuel between a reactor building and an auxiliary building in a pressure tube reactor or the like. Structure: The structure comprises an oblique transfer chute loosely penetrating the reactor building, reactor container and auxiliary building, a transfer chute support outer cylinder surrounding the transfer chute and having one end coupled to the transfer chute and other end coupled to the container, flexible seal members respectively provided on the reactor building side and on the auxiliary building side and surrounding the transfer chute and a slidable support supported on the side of the auxiliary building such that it can be in frictional contact with the outer periphery of the transfer chute. With this construction, the relative displacements of various parts caused by an earthquake or the like can be absorbed by the support outer cylinder, flexible seals and slidable support. (Ikeda, J.)

Safety equipment in a reactor

International Nuclear Information System (INIS)

Shiratori, Hirozo; Ishiyama, Satoshi; Ugawa, Yukio.

1976-01-01

Object: To safely retain, even if fuel should be molten and flown through the bottom of a container in a reactor, the molten fuel to remove heat generation of the fuel to prevent occurrence of a critical trouble. Structure: A reactor container housing a core and coolant has thereunder a separation dome in a central portion thereof and a partitioning plate coaxially and circularly disposed in the periphery of the separation dome, with a tray formed of magnesium oxide being disposed. Further, a cooling path system is provided so as to surround the tray. The cooling path system and the reactor container are surrounded and protected by a reactor wall provided with heat insulating refractory bricks, a coolant pouring system extends through the reactor wall, and the coolant is supplied to the tray. (Furukawa, Y.)

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.)

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

International Nuclear Information System (INIS)

Baratta, A.J.

1997-01-01

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

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

Energy Technology Data Exchange (ETDEWEB)

Baratta, A.J.

1997-07-01

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

Nuclear reactor cavity streaming shield

International Nuclear Information System (INIS)

Klotz, R.J.; Stephen, D.W.

1978-01-01

The upper portion of a nuclear reactor vessel supported in a concrete reactor cavity has a structure mounted below the top of the vessel between the outer vessel wall and the reactor cavity wall which contains hydrogenous material which will attenuate radiation streaming upward between vessel and the reactor cavity wall while preventing pressure buildup during a loss of coolant accident

Structure optimization of CFB reactor for moderate temperature FGD

Energy Technology Data Exchange (ETDEWEB)

Li, Yuan; Zhang, Jie; Zheng, Kai; You, Changfu [Tsinghua Univ., Beijing (China). Dept. of Thermal Engineering; Ministry of Education, Beijing (China). Key Lab. for Thermal Science and Power Engineering

2013-07-01

The gas velocity distribution, sorbent particle concentration distribution and particle residence time in circulating fluidized bed (CFB) reactors for moderate temperature flue gas desulfurization (FGD) have significant influence on the desulfurization efficiency and the sorbent calcium conversion ratio for sulfur reaction. Experimental and numerical methods were used to investigate the influence of the key reactor structures, including the reactor outlet structure, internal structure, feed port and circulating port, on the gas velocity distribution, sorbent particle concentration distribution and particle residence time. Experimental results showed that the desulfurization efficiency increased 5-10% when the internal structure was added in the CFB reactor. Numerical analysis results showed that the particle residence time of the feed particles with the average diameter of 89 and 9 {mu}m increased 40% and 17% respectively, and the particle residence time of the circulating particles with the average diameter of 116 {mu}m increased 28% after reactor structure optimization. The particle concentration distribution also improved significantly, which was good for improving the contact efficiency between the sorbent particles and SO{sub 2}. In addition, the optimization guidelines were proposed to further increase the desulfurization efficiency and the sorbent calcium conversion ratio.

Evaluation of a hydrogen sensor for nuclear reactor containment monitoring

International Nuclear Information System (INIS)

Hoffheins, B.S.; McKnight, T.E.; Lauf, R.J.; Smith, R.R.; James, R.E.

1997-01-01

Measurement of hydrogen concentration in containment atmospheres in nuclear plants is a key safety capability. Current technologies require extensive sampling systems and subsequent maintenance and calibration costs can be very expensive. A new hydrogen sensor has been developed that is small and potentially inexpensive to install and maintain. Its size and low power requirement make it suitable in distributed systems for pinpointing hydrogen buildup. This paper will address the first phase of a testing program conducted to evaluate this sensor for operation in reactor containments

Strategy for 100-year life of the ACR-1000 concrete containment structure

International Nuclear Information System (INIS)

Abrishami, H.; Elgohary, M.

2006-01-01

The purpose of this paper is to present the Plant Life Management (PLiM) strategy for the concrete containment structure of the ACR-1000 (Advanced CANDU Reactor) designed by AECL. The ACR-1000 is designed for 100-year plant life including 60-year operating life and additional 40-year decommissioning period of time. The approach adopted for the PLiM strategy of the concrete containment structure is a preventive one, key areas being: 1) design methodology, 2) material performance and 3) life cycle management and ageing management program. In the design phase, in addition to strength and serviceability, durability is a major requirement during the service life and decommissioning phase of the ACR structure. Parameters affecting durability design include: a) concrete performance, b) structural application, and c) environmental conditions. Due to the complex nature of the environmental effects acting on structures during the service life of project, it is considered that true improved performance during the service life can be achieved by improving the material characteristics. Many recent innovations in advanced concrete materials technology have made it possible to produce modern concrete such as high-performance concrete with exceptional performance characteristics. In this paper, the PLiM strategy for the ACR-1000 concrete containment is presented. In addition to addressing the design methodology and material performance areas, a systematic approach for ageing management program for the concrete containment structure is presented. (author)

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.

Decommissioning of the secondary containment of the steam generating heavy water reactor at UKAEA-Winfrith

International Nuclear Information System (INIS)

Miller, Keith; Cornell, Rowland; Parkinson, Steve; McIntyre, Kevin; Staples, Andy

2007-01-01

Available in abstract form only. Full text of publication follows: The Winfrith SGHWR was a prototype nuclear power plant operated for 23 years by the United Kingdom Atomic Energy Authority (UKAEA) until 1990 when it was shut down permanently. The current Stage 1 decommissioning contract is part of a multi-stage strategy. It involves the removal of all the ancillary plant and equipment in the secondary containment and non-containment areas ahead of a series of contracts for the decommissioning of the primary containment, the reactor core and demolition of the building and all remaining facilities. As an outcome of a competitive tending process, the Stage 1 decommissioning contract was awarded to NUKEM with operations commencing in April 2005. The decommissioning processes involved with these plant items will be described with some emphasis of the establishment of multiple work-fronts for the production, recovery, treatment and disposal of mainly tritium-contaminated waste arising from its contact with the direct cycle reactor coolant. The means of size reduction of a variety of large, heavy and complex items of plant made from a range of materials will also be described with some emphasis on the control of fumes during hot cutting operations and establishing effective containments within a larger secondary containment structure. Disposal of these wastes in a timely and cost-effective manner is a major challenge facing the decommissioning team and has required the development of a highly efficient means of packing the resultant materials into mainly one-third height ISO containers for disposal as LLW. Details of the quantities of LLW and exempt wastes handled during this process will be given with a commentary about the difficulty in segregating these two waste streams efficiently. (authors)

Theoretical investigations to the mass- and energy transport in light water reactor containment buildings

International Nuclear Information System (INIS)

Hocke, K.D.; Bisanz, R.; Ploeger, G.

1982-12-01

In this report loss-of-coolant accidents in water-cooled nuclear power reactors will be described with regard to their effects on the reactor containment system. The general presentation of containment systems and the accident sequences is completed by performances of the physical and mathematical treatment of the flow dynamics. Following a survey of existing computer codes for containment analysis problems, a detailed model description of the computer codes ZOCO 6 and BEACON-MOD 2A is given. A variation of important model parameters within the ZOCO 6 program explains the field of application and leads to a valuation of the model assumptions. The obtained results are presented graphically including latest publications. (orig.) [de

Design of the containment structure in prestressed concrete for the Embalse-Cordoba Nuclear Power Plant

International Nuclear Information System (INIS)

Godoy, A.R.; Marinelli, C.A.; Gruenbaum, C.E.

1978-01-01

The design of a typical prestressed concrete containment structure for a 600 MW Candu - PHW Reactor, presently under construction at Embalse - Cordoba, Argentina is briefly described. The structural behaviour , adcpted prestressing system and tendon pattern are described. Afterwards the evaluation of the prestressing forces as well as the losses assessment and the prestressing sequence are discussed. Finally, some conclusions are drawn in the light of the experience gained at different stages of the construction. (Author)

Deposition of RuO{sub 4} on various surfaces in a nuclear reactor containment

Energy Technology Data Exchange (ETDEWEB)

Holm, Joachim, E-mail: joachim.holm@chalmers.s [Department of Nuclear Chemistry, Chalmers University of Technology, Se-412 96 Gothenburg (Sweden); Glaenneskog, Henrik [Ringhals AB, SE-430 22, Vaeroebacka (Sweden); Ekberg, Christian [Department of Nuclear Chemistry, Chalmers University of Technology, Se-412 96 Gothenburg (Sweden)

2009-07-01

During a severe nuclear reactor accident with air ingress, ruthenium can be released from the nuclear fuel in the form of ruthenium tetroxide. Hence, it is important to investigate how the reactor containment is able to reduce the source term of ruthenium. The aim of this work was to investigate the deposition of gaseous ruthenium tetroxide on aluminium, copper and zinc, which all appear in relatively large amounts in reactor containment. The experiments show that ruthenium tetroxide is deposited on all the metal surfaces, especially on the copper and zinc surfaces. A large deposition of ruthenium tetroxide also appeared on the relatively inert glass surfaces in the experimental set-ups. The analyses of the different surfaces, with several analytical methods, showed that the form of deposited ruthenium was mainly ruthenium dioxide.

Tritium concentration reducing method in atmosphere in nuclear reactor containment facility

International Nuclear Information System (INIS)

Hirasawa, Yoshiya; Kigoshi, Yasutane; Yonenaga, Haruo.

1992-01-01

A portion of water content in an atmosphere is condensed by a condensation/evaporation device disposed in a nuclear reactor containment building and then a portion of the condensed water is evaporated in the atmosphere. A portion of hydrogen nuclides constituting the evaporated water content is subjected to isotopic exchange with tritium nuclides in the atmosphere. A portion of water content in the atmosphere applied with the isotopic exchange is condensed in the condensation/evaporation device. That is, the hydrogen nuclides in steams are applied with isotopic exchange with tritium nuclides, and steams incorporating tritium nuclides are condensed again in the condensation/evaporation device, to transfer the tritium nuclides in the atmosphere to condensed water. The condensed water is recovered without releasing the tritium nuclides to the outside of the reactor containment building, thereby enabling to reduce the tritium concentration in the atmosphere. (N.H.)

Electrochemical aspects on corrosion in Swedish reactor containments; Elektrokemiska aspekter paa korrosion i svenska reaktorinneslutningar

Energy Technology Data Exchange (ETDEWEB)

Ullberg, Mats [Studsvik Nuclear AB, SE-611 82 Nykoeping (Sweden)

2006-10-15

Post-stressed concrete is used in all Swedish nuclear reactor containments. Steel in concrete is normally protected from corrosion by the highly alkaline pore solution in concrete. A passive film develops on the surface of steel in contact with the pore solution. However, corrosion may still occur under special circumstances. It is therefore desirable to monitor the corrosion status of the containment. A review of the corrosion experience with steel in concrete strongly suggests that the potential problem of most concern for the Swedish reactor containments is cavity formation during grouting of tendons and of penetrations in the containment wall. Cavities break the contact between alkaline grout and steel. Corrosion is then possible, provided the relative humidity is high enough. Normal methods for inspection of the corrosion status of steel reinforcement in concrete are not applicable to very heavy structures like reactor containments. Since inspections are difficult to carry out, it is important that they be focused on the most susceptible portions of the containment. This report is an attempt to assemble potentially useful background information. The original intention was to focus on electrochemical methods of investigation. When it was realized that the potential use of electrochemical methods was limited, the scope of the review was broadened. The present as well as previous investigations indicate that nondestructive testing of grouted tendons is the outstanding problem in the condition assessment of Swedish nuclear reactor containments. Grouted tendons are also used in a very large number of bridges built since the early 1950s. The experience gained in connection with bridges has therefore been investigated. The need for a testing method for grouted tendons in bridges has long been strongly felt and development work has been in progress since the early 1970-ies, for example within the Strategic Highway Research Project in the Unite States. Potential

Transients in a reactor containment after a HCDA

International Nuclear Information System (INIS)

Ghosh, A.K.

1984-01-01

The consequences of a hypothetical core meltdown accident in a fast reactor is analysed. Shock waves are generated in the surrounding medium after the energy release, which is assumed to be instantaneous and at a point. After discussing the difference in the predicted and experimentally observed peak pressure a semi-empirical approach is taken to arrive at a better estimate. This defines the loading on the containment, which is idealised as a combination of a shell and a plate. To simplify the analysis the shell is assumed to be in plane strain and axial symmetry is assumed for both components. The shell response is evaluated by using elasticity theory. Numerical results are presented for peak overpressure and pressure-time history and dynamic stresses in the containment. (Author) [pt

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

Pressurized Water Reactor containment in Russia

International Nuclear Information System (INIS)

Taymouri, Majid.

1993-01-01

One of the most important systems of nuclear power plants from an economical point of view and view point of safety is containment; Therefore, the containments designed in Russia were studied in the first chapter. Russian general rules and requirements of structure of accident localization system were illustrated. Methods of accident localization system rooms tested for tightness and strength are presented in chapter three. Russian specialists have been working hard to ensure the safety culture in building structures and operational procedures and the have successfully implemented these objectives in new nuclear power plant designs and rules

Catalyst support structure, catalyst including the structure, reactor including a catalyst, and methods of forming same

Science.gov (United States)

Van Norman, Staci A.; Aston, Victoria J.; Weimer, Alan W.

2017-05-09

Structures, catalysts, and reactors suitable for use for a variety of applications, including gas-to-liquid and coal-to-liquid processes and methods of forming the structures, catalysts, and reactors are disclosed. The catalyst material can be deposited onto an inner wall of a microtubular reactor and/or onto porous tungsten support structures using atomic layer deposition techniques.

Passive heat removal from containment

International Nuclear Information System (INIS)

Gou, P.F.; Townsend, H.E.

1990-01-01

This patent describes a heat removal system for removing heat from a containment of a nuclear reactor. It comprises: a sealed suppression chamber in the containment; means for venting steam from the nuclear reactor into the suppression chamber upon occurrence of an event requiring dissipation of heat from the nuclear reactor. The suppression chamber containing a quantity of water; the suppression chamber having a gas-containing space above the water; a heat exchanger disposed within the gas-containing space of the suppression chamber; the heat exchanger including an enclosed structure for holding a heat-exchange fluid; means for metering a supply of heat-exchange fluid to the heat exchanger to maintain a predetermined level thereof in the enclosed structure. The heat-exchange fluid boiling in the heat exchanger in consequence of heat transfer thereto from steam present in the suppression chamber; means for separating a heat-exchange fluid vapor in the heat exchanger from the heat-exchange fluid; and means for discharging the vapor immediately following its separation from heat-exchange fluid directly from the heat exchanger to a location exterior of the containment, whereby heat is discharged from the suppression chamber, and the containment is maintained at a temperature and pressure below its design value

Packed bed reactor for degradation of simulated cyanide-containing wastewater.

Science.gov (United States)

Kumar, Virender; Kumar, Vijay; Bhalla, Tek Chand

2015-10-01

The discharge of cyanide-containing effluents into the environment contaminates water bodies and soil. Effective methods of treatment which can detoxify cyanide are the need of the hour. The aim of the present study is to develop a bioreactor for complete degradation of cyanide using immobilized cells of Serratia marcescens RL2b. Alginate-entrapped cells of S. marcescens RL2b were used for complete degradation of cyanide in a packed bed reactor (PBR). Cells grown in minimal salt medium (pH 6.0) were harvested after 20 h and exhibited 0.4 U mg -1  dcw activity and 99 % cyanide degradation in 10 h. These resting cells were entrapped using 3 % alginate beads and packed in a column reactor (20 × 1.7 cm). Simulated cyanide (12 mmol l -1 )-containing wastewater was loaded and fractions were collected after different time intervals at various flow rates. Complete degradation of 12 m mmol l -1 (780 mg l -1 ) cyanide in 10 h was observed at a flow rate of 1.5 ml h -1 . The degradation of cyanide in PBR showed direct dependence on retention time. The retention time of cyanide in the reactor was 9.27 h. The PBR can degrade 1.2 g of cyanide completely in 1 day.

Nuclear reactor

International Nuclear Information System (INIS)

Hattori, Sadao; Sato, Morihiko.

1994-01-01

Liquid metals such as liquid metal sodium are filled in a reactor container as primary coolants. A plurality of reactor core containers are disposed in a row in the circumferential direction along with the inner circumferential wall of the reactor container. One or a plurality of intermediate coolers are disposed at the inside of an annular row of the reactor core containers. A reactor core constituted with fuel rods and control rods (module reactor core) is contained at the inside of each of the reactor core containers. Each of the intermediate coolers comprises a cylindrical intermediate cooling vessels. The intermediate cooling vessel comprises an intermediate heat exchanger for heat exchange of primary coolants and secondary coolants and recycling pumps for compulsorily recycling primary coolants at the inside thereof. Since a plurality of reactor core containers are thus assembled, a great reactor power can be attained. Further, the module reactor core contained in one reactor core vessel may be small sized, to facilitate the control for the reactor core operation. (I.N.)

A method to calculate flux distribution in reactor systems containing materials with grain structure

International Nuclear Information System (INIS)

Stepanek, J.

1980-01-01

A method is proposed to compute the neutron flux spatial distribution in slab, spherical or cylindrical systems containing zones with close grain structure of material. Several different types of equally distributed particles embedded in the matrix material are allowed in one or more zones. The multi-energy group structure of the flux is considered. The collision probability method is used to compute the fluxes in the grains and in an ''effective'' part of the matrix material. Then the overall structure of the flux distribution in the zones with homogenized materials is determined using the DPN ''surface flux'' method. Both computations are connected using the balance equation during the outer iterations. The proposed method is written in the code SURCU-DH. Two testcases are computed and discussed. One testcase is the computation of the eigenvalue in simplified slab geometry of an LWR container of one zone with boral grains equally distributed in an aluminium matrix. The second is the computation of the eigenvalue in spherical geometry of the HTR pebble-bed cell with spherical particles embedded in a graphite matrix. The results are compared to those obtained by repeated use of the WIMS Code. (author)

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

Energy Technology Data Exchange (ETDEWEB)

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

2008-12-15

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

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

International Nuclear Information System (INIS)

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

2008-01-01

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

Structural materials issues for the next generation fission reactors

Science.gov (United States)

Chant, I.; Murty, K. L.

2010-09-01

Generation-IV reactor design concepts envisioned thus far cater to a common goal of providing safer, longer lasting, proliferation-resistant, and economically viable nuclear power plants. The foremost consideration in the successful development and deployment of Gen-W reactor systems is the performance and reliability issues involving structural materials for both in-core and out-of-core applications. The structural materials need to endure much higher temperatures, higher neutron doses, and extremely corrosive environments, which are beyond the experience of the current nuclear power plants. Materials under active consideration for use in different reactor components include various ferritic/martensitic steels, austenitic stainless steels, nickel-base superalloys, ceramics, composites, etc. This article addresses the material requirements for these advanced fission reactor types, specifically addressing structural materials issues depending on the specific application areas.

Dynamic analysis of reactor containment subjected to aircraft impact loading

International Nuclear Information System (INIS)

Li Xiaotian; He Shuyan

2004-01-01

In this paper, dynamic character of reactor containment subjected to aircraft impact loading is analyzed with MSC.DYTRAN program. The displacement of concrete and velocity curve of airplane is obtained. The results of the different material model are compared with empirical formula. It is concluded that reasonable result can be obtained using cap model for concrete

Nuclear power plant containment construction

International Nuclear Information System (INIS)

Schabert, H.P.; Danisch, R.; Strickroth, E.

1975-01-01

The Nuclear Power Plant Containment Construction includes the spherical steel safety enclosure for the reactor and the equipment associated with the reactor and requiring this type of enclosure. This steel enclosure is externally structurally protected against accident by a concrete construction providing a foundation for the steel enclosure and having a cylindrical wall and a hemispherical dome, these parts being dimensioned to form an annular space surrounding the spherical steel enclosure, the latter and the concrete construction heretofore being concentrically arranged with respect to each other. In the disclosed construction the two parts are arranged with their vertical axis horizontally offset from each other so that opposite to the offsetting direction of the concrete construction a relatively large space is formed in the now asymmetrical annular space in which reactor auxiliary equipment not requiring enclosure by the steel containment vessel or safety enclosure, may be located outside of the steel containment vessel and inside of the concrete construction where it is structurally protected by the latter

Vibration test on KMRR reactor structure and primary cooling system piping

Energy Technology Data Exchange (ETDEWEB)

Chung, Seung Hoh; Kim, Tae Ryong; Park, Jin Hoh; Park, Jin Suk; Ryoo, Jung Soo [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

1994-10-01

Most equipments, piping systems and reactor structures in nuclear power plants are subjected to flow induced vibration due to high temperature and high pressure coolant flowing inside or outside of the equipments, systems and structures. Because the flow induced vibration sometimes causes significant damage to reactor structures and piping systems, it is important and necessary to evaluate the vibration effect on them and to prove their structural integrity. Korea Multipurpose Research Reactor (KMRR) being constructed by KAERI is 30 MWt pool type research reactor. Since its main structures and piping systems were designed and manufactured in accordance with the standards and guidelines for commercial nuclear power plant, it was decided to evaluate their vibratory response in accordance with the standards and guidelines for commercial NPP. The objective of this vibration test is the assessment of vibration levels of KMRR reactor structure and primary cooling piping system for their structural integrity under the steady-state or transient operating condition. 38 figs, 14 tabs, 2 refs. (Author).

Vibration test on KMRR reactor structure and primary cooling system piping

International Nuclear Information System (INIS)

Chung, Seung Hoh; Kim, Tae Ryong; Park, Jin Hoh; Park, Jin Suk; Ryoo, Jung Soo

1994-10-01

Most equipments, piping systems and reactor structures in nuclear power plants are subjected to flow induced vibration due to high temperature and high pressure coolant flowing inside or outside of the equipments, systems and structures. Because the flow induced vibration sometimes causes significant damage to reactor structures and piping systems, it is important and necessary to evaluate the vibration effect on them and to prove their structural integrity. Korea Multipurpose Research Reactor (KMRR) being constructed by KAERI is 30 MWt pool type research reactor. Since its main structures and piping systems were designed and manufactured in accordance with the standards and guidelines for commercial nuclear power plant, it was decided to evaluate their vibratory response in accordance with the standards and guidelines for commercial NPP. The objective of this vibration test is the assessment of vibration levels of KMRR reactor structure and primary cooling piping system for their structural integrity under the steady-state or transient operating condition. 38 figs, 14 tabs, 2 refs. (Author)

Marine reactor pressure vessels dumped in the Kara Sea

International Nuclear Information System (INIS)

Mount, M.E.

1997-01-01

Between 1965 and 1988, 16 marine reactors from seven Russian submarines and the icebreaker Lenin, each of which suffered some form of reactor accident, were dumped in a variety of containments, using a number of sealing methods, at five sites in the Kara Sea. All reactors were dumped at sites that varied in depth from 12 to 300 m and six contained their spent nuclear fuel (SNF). This paper examines the breakdown of the reactor pressure vessel (RPV) barriers due to corrosion, with specific emphasis on those RPVs containing SNF. Included are discussions of the structural aspects of the steam generating installations and their associated RPVs, a summary of the disposal operations, assumptions on corrosion rates of structural and filler materials, and an estimate of the structural integrity of the RPVs at the present time (1996) and in the year 2015

Conceptual design of nuclear fusion power reactor DREAM. Reactor structures and remote maintenance

International Nuclear Information System (INIS)

Nishio, Satoshi; Seki, Yasushi; Ueda, Shuzo; Kurihara, Ryoichi; Adachi, Junichi; Yamazaki, Seiichiro; Hashimoto, Toshiyuki.

1997-01-01

Nuclear fusion reactors are required to be able to compete another energy sources in economy, reliability, safety and environmental integrity for commercial use. In the DREAM (DRastically EAsy Maintenance) reactor, a very low activated material of SiC/SiC composite has been introduced for the structural material, a reactor configuration for very easy maintenance and the helium gas of a high temperature for the cooling system, and hence DREAM has been proven to be very attractively as the commercial power reactor due to the high availability and efficiency of the plant and minimization of radioactive wastes. (author)

High-Temperature Gas-cooled Reactor steam-cycle/cogeneration lead plant reactor vessel: system design description

International Nuclear Information System (INIS)

1983-01-01

The Reactor Vessel System contains the primary coolant inventory within a gas-tight pressure boundary, and provides the necessary flow paths and overpressure protection for this pressure boundary. The Reactor Vessel System also houses the components of the Reactor System, the Heat Transport System, and the Auxiliary Heat Removal System. The scope of the Reactor Vessel System includes the prestressed concrete reactor vessel (PCRV) structure with its reinforcing steel and prestressing components; liners, penetrations, closures, and cooling water tubes attached to the concrete side of the liner; the thermal barrier (insulation) on the primary coolant side of the liner; instrumentation for structural monitoring; and a pressure relief system. Specifications are presented

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)

Structure of thermonuclear reactor wall

International Nuclear Information System (INIS)

Yamazaki, Seiichiro.

1991-01-01

In a thermonuclear reactor wall, there has been a worry that the brazing material is melted by high temperature heat and particle load, to peel off the joined portion and the protecting material is destroyed by temperature elevation, to expose the heat sink material. Then, in the reactor core structures of a thermonuclear reactor, such as a divertor plate comprising a protecting material made of carbon material and the heat sink material joined by brazing, a plate material made of a so-called refractory metal having a high atomic number such as tungsten, molybdenum or the alloy thereof is embedded or attached to an accurate position of the protecting material. This can prevent the brazing portion from destruction by escaping electrons generated upon occurrence of abnormality in the thermonuclear reactor, and peeling or destroy of the protecting material and the heat sink material. Sufficient characteristics of plasmas can always be maintained by disposing a material having a small atomic number, for example, carbon material, to the position facing to the plasmas. (N.H.)

Space Fission Reactor Structural Materials: Choices Past, Present and Future

International Nuclear Information System (INIS)

Busby, Jeremy T.; Leonard, Keith J.

2007-01-01

Nuclear powered spacecraft will enable missions well beyond the capabilities of current chemical, radioisotope thermal generator and solar technologies. The use of fission reactors for space applications has been considered for over 50 years, although, structural material performance has often limited the potential performance of space reactors. Space fission reactors are an extremely harsh environment for structural materials with high temperatures, high neutron fields, potential contact with liquid metals, and the need for up to 15-20 year reliability with no inspection or preventative maintenance. Many different materials have been proposed as structural materials. While all materials meet many of the requirements for space reactor service, none satisfy all of them. However, continued development and testing may resolve these issues and provide qualified materials for space fission reactors.

Organic matter and containment of uranium and fissiogenic isotopes at the Oklo natural reactors

International Nuclear Information System (INIS)

Nagy, B.; Rigali, M.J.; Davis, D.W.; Parnell, J.

1991-01-01

Some of the Precambrian natural fission reactors at Oklo in Gabon contain abundant organic matter, part of which was liquefied at the time of criticality and subsequently converted to a graphitic solid. The liquid organic matter helps to reduce U(VI) to U(IV) from aqueous solutions, resulting in the precipitation of uraninite. It is known that in the prevailing reactor environments, precipitated uraninite grains incorporated fission products. We report here observations which show that these uraninite crystals were held immobile within the re-solidified, graphitic bituminous organics at Oklo thus enhanced radionuclide containment. Uraninite encased in solid graphitic matter in the organic-rich reactor zones lost virtually no fissiogenic lanthanide isotopes. The first major episode of uranium and lead migration was caused by the intrusion of a swarm of adjacent dolerite dykes about 1,100 Myr after the reactors went critical. Our results from Oklo imply that the use of organic, hydrophobic solids such as graphitic bitumen as a means of immobilizing radionuclides in pre-treated nuclear waste warrants further investigation. (author)

Aircraft Impact Assessment of APR1400 Reactor Containment Building

International Nuclear Information System (INIS)

Moon, Il Hwan; Kim, Do Yeon; Kim, Jae Hee; Kim, Sang Yun

2011-01-01

The implementation of a protection to withstand aircraft impact on safety-related structures and systems is basically based on a probabilistic evaluation for each site, if the licensing body doesn't require a deterministic approach. Existing nuclear power plants in Korea were designed based on the probabilistic approach, and the aircraft impact hazard remained less than a probability of 10 -7 . However, a man-made aircraft impact have been considered as a possible external accident for the nuclear power plant. New plant designs that are to be constructed in the U.S. after July 2009 must consider the effect of impact from a large commercial aircraft according to the requirements of 10 CFR 50.150. Especially, Reactor Containment Building (RCB) housing the safety-related equipment and fuels should be protected safely against aircraft crash without perforation and scabbing failure of external wall. APR1400 RCB is constructed as a prestressed concrete containment vessel (PCCV) which is surrounded by the auxiliary building housing additional safety-related equipment and other systems. In this study, the aircraft impact analyses for the RCB are carried out using Riera forcing function and aircraft model. Considered external wall thickness is 4 ft 6 in. for the cylindrical wall and 4 ft for the dome. Actual strengths of concrete and steel are considered as the material properties. For these analyses, the dynamic increment factor and concrete aging effect are considered in accordance with NEI 07-13(2011)

Advances in the analysis and design of concrete structures, metal containments and liner plate for extreme loads

International Nuclear Information System (INIS)

Stevenson, J.D.; Eibl, J.; Curbach, M.; Johnson, T.E.; Daye, M.A.; Riera, J.D.; Nemet, J.; Iyengar, K.T.S.

1992-01-01

The material presented in this paper summarizes the progress that has been made in the analysis, design, and testing of concrete structures. The material is summarized in the following documents: Part I: Containment Design Criteria and Loading Combinations; Part II: Reinforced and Prestressed Concrete Behavior; Part III: Concrete Containment Analysis, Design and Related Testing; Part IV: Impact and Impulse Loading and Response Prediction; Part V: Metal Containments and Liner Plate Systems; Part VI: Prestressed Reactor Vessel Design, Testing and Analysis. (orig.)

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

Reactor structure and superconducting magnet system of ITER

International Nuclear Information System (INIS)

Tada, Eisuke; Yoshida, Kiyoshi; Shibanuma, Kiyoshi; Okuno, Kiyoshi; Tsuji, Hiroshi; Shimamoto, Susumu

1993-01-01

Fusion Experimental Reactors are one of the major steps toward realization of the fusion energy and the key objective are to demonstrate the scientific and technological feasibility prior to the Demo Fusion Reactor. ITER (International Thermonuclear Experimental Reactor) is one of experimental reactors and the conceptual design has been completed by the united efforts of USA, USSR, EC and Japan. In parallel with the conceptual design, key technology development in various areas has being conducted. This paper describes the overall design concepts and the latest technological achievements of the ITER reactor structure and superconducting magnet system. (author)

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

U.S. fast reactor materials and structures program

International Nuclear Information System (INIS)

Harms, W.O.; Purdy, C.M.

1984-01-01

The U.S. DOE has sponsored a vigorous breeder reactor materials and structures program for 15 years. Important contributions have resulted from this effort in the areas of design (inelastic rules, verified methods, seismic criteria, mechanical properties data); resolution of licensing issues (technical witnessing, confirmatory testing); construction (fabrication/welding procedures, nondestructive testing techniques); and operation (sodium purification, instrumentation and chemical analysis, radioactivity control, and in-service inspection. The national LMFBR program currently is being restructured. The Materials and Structures Program will focus its efforts in the following areas: (1) removal of anticipated licensing impediments through confirmation of the adequacy of structural design methods and criteria for components containing welds and geometric discontinuities, the generation of mechanical properties for stainless steel castings and weldments, and the evaluation of irradiation effects; (2) qualification of modified 9 Cr-1 Mo steel and tribological coatings for design flexibility; (3) development of improved inelastic design guidelines and procedures; (4) reform of design codes and standards and engineering practices, leading to simpler, less conservative rules and to simplified design analysis methods; and (5) incorporation of information from foreign program

Structural integrity aspects of reactor safety

Indian Academy of Sciences (India)

A large experimental programme supported the structural integrity demonstration. ... Categories in which the structures, systems and components (SSC) are .... One of the ways in which the decision to live with the defect can be aided is the .... The Advanced Heavy Water Reactor (AHWR) (figure 18) being designed by BARC ...

Analysis of design strategies for mitigating the consequences of lithium fire within containment of controlled thermonuclear reactors

Energy Technology Data Exchange (ETDEWEB)

Dube, D A; Kazimi, M S

1978-07-01

A lithium combustion model (LITFIRE) was developed to describe the physical and chemical processes which occur during a hypothetical lithium spill and fire. The model was used to study the effectiveness of various design strategies for mitigating the consequences of lithium fire, using the UWMAK-III features as a reference design. Calculations show that without any special fire protection measures, the containment may reach pressures of up to 32 psig when one coolant loop is spilled inside the reactor building. Temperatures as high as 2000/sup 0/F would also be experienced by some of the containment structures. These consequences were found to diminish greatly by the incorporation of a number of design strategies including initially subatmospheric containment pressures, enhanced structural surface heat removal capability, initially low oxygen concentrations, and active post-accident cooling of the containment gas. The EBTR modular design was found to limit the consequences of a lithium spill, and hence offers a potential safety advantage. Calculations of the maximum flame temperature resulting from lithium fire indicate that none of the radioactive first wall materials under consideration would vaporize, and only a few could possibly melt.

Analysis of design strategies for mitigating the consequences of lithium fire within containment of controlled thermonuclear reactors

International Nuclear Information System (INIS)

Dube, D.A.; Kazimi, M.S.

1978-07-01

A lithium combustion model (LITFIRE) was developed to describe the physical and chemical processes which occur during a hypothetical lithium spill and fire. The model was used to study the effectiveness of various design strategies for mitigating the consequences of lithium fire, using the UWMAK-III features as a reference design. Calculations show that without any special fire protection measures, the containment may reach pressures of up to 32 psig when one coolant loop is spilled inside the reactor building. Temperatures as high as 2000 0 F would also be experienced by some of the containment structures. These consequences were found to diminish greatly by the incorporation of a number of design strategies including initially subatmospheric containment pressures, enhanced structural surface heat removal capability, initially low oxygen concentrations, and active post-accident cooling of the containment gas. The EBTR modular design was found to limit the consequences of a lithium spill, and hence offers a potential safety advantage. Calculations of the maximum flame temperature resulting from lithium fire indicate that none of the radioactive first wall materials under consideration would vaporize, and only a few could possibly melt

Structural Integrity Evaluation of the KALIMER-600 Reactor Core Support Structure

International Nuclear Information System (INIS)

Park, Chang Gyu; Kim, Jong Bum; Lee, Jae Han

2005-01-01

KALIMER-600(Korea Advanced LIquid MEtal Reactor, 600MWe) is a pool type sodium-cooled liquid metal reactor. Since the normal operating temperature of KALIMER-600 is 545 .deg. C, the reactor structures in the hot pool region are designed and evaluated according to the elevated temperature design rules such as the ASME Boiler and Pressure Vessel Code Section III, Subsection NH. Since the core support structure of KALIMER-600 is in the cold pool region under 400 .deg. C, a high temperature inelastic behavior is not expected. Thus the stress and fatigue limits are the main concerns to assure the structural design integrity following the ASME Subsection NG. In this paper, the evaluations of the stress and fatigue damage for the core support structure of KALIMER-600 are carrried out in the case of a normal operation condition using the rules of ASME Subsection NG. To obtain the stress values, a heat transfer analysis and a stress analysis under a combined loading condition are performed. From the stress distribution results, the critical sections are selected and the stress and fatigue limits are evaluated for the selected regions

FBR type reactors

International Nuclear Information System (INIS)

Nakamura, Tsugio.

1986-01-01

Purpose: To ensure the thermal integrity of a reactor vessel in FBR type reactors by preventing sodium vapors or the likes from intruding into a shielding chamber and avoiding spontaneous convection thereof. Constitution: There are provided a shielding plug for shielding the upper opening of a reactor container, an annular thermal member disposed to the circumferential side in the container, a shielding member for shielding upper end of the shielding chamber and a plurality of convection preventive plates suspended from the thermal member into the shielding chamber, and the shielding chamber is communicated by way of the relatively low temperature portion of the container with a gas communication pipe. That is, by closing the upper end of the shielding chamber with the shielding member, coolant vapors, etc. can be prevented from intruding into the shielding chamber. Further, the convection preventive plates prevent the occurrence of spontaneous convection in the shielding chamber. Further, the gas communication pipe absorbs the expansion and contraction of gases in the shielding chamber to effectively prevent the deformation or the like for each of the structural materials. In this way, the thermal integrity of the reactor container can surely be maintained. (Horiuchi, T.)

Earthquake-proof supporting structure in reactor vessel

International Nuclear Information System (INIS)

Sakurai, Akio; Sekine, Katsuhisa; Madokoro, Manabu; Katoono, Shin-ichi; Konno, Mutsuo; Suzuki, Takuro.

1990-01-01

Conventional earthquake-proof structure comprises a vessel vibration stopper integrated to a reactor vessel, powder for restricting the horizontal displacements, a safety vessel surrounds the outer periphery of the reactor vessel and a safety vessel vibration stopper integrated therewith, which are fixed to buildings. However, there was a problem that a great amount of stresses are generated in the base of the reactor vessel vibration stopper due to reaction of the powders which restrict thermal expansion. In order to remarkably reduce the reaction of the powers, powders are charged into a spaces formed between each of the reactor vessel vibration stopper, the safety vessel vibration stopper and the flexible member disposed between them. According to this constitution, the reactor vessel vibration stopper does not undergo a great reaction of the powers upon thermal expansion of the reactor vessel to moderate the generated stresses, maintain the strength and provide earthquake-proof supporting function. (N.H.)

Structure of steel reactor building and construction method therefor

International Nuclear Information System (INIS)

Yamakawa, Toshikimi.

1997-01-01

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

Analysis of a possible experimental assessment of a prototype fuel element containing burnable poison in the RA-3 reactor

International Nuclear Information System (INIS)

Lerner, Ana Maria; Madariaga, Marcelo

2002-01-01

The Argentine RA-3 research reactor (5 MW) is presently operated with LEU fuel by the National Atomic Energy Commission (CNEA). It belongs to the group of nuclear installations controlled, from the radiological and nuclear safety point of view, by the Nuclear Regulatory Authority (ARN). A new type of fuel elements containing burnable absorbers, with similar enrichment as the standard fuel elements but greater fissile contents, has recently been proposed for a new Argentine reactor design (RRR). In this framework the ARN considers interesting, if technically possible, the performance of an experiment in the RA-3 reactor. The experiment might enable, for such fuel element containing burnable poison, the verification of its neutronic behaviour under irradiation as well as a validation of the calculation line by comparison to measured values. It should be desirable that such experiment could reproduce as much as possible those conditions estimated for the RRR reactor, still under design in Argentina, having Silicide fuel elements with burnable poison, in the shape of cadmium wires in their structure. We here analyse a possible experiment consisting in the loading of a prototype fuel element with burnable poison in a normally loaded RA-3 core configuration. It would essentially be a standard RA-3 fuel element, having cadmium wires in its frame. This experiment would enable the verification of the prototype behaviour under irradiation, its operation limits and conditions, and particularly, the reactivity safety margins established in Argentine Standards, both calculated and measured. The main part of the experiment would imply some 200 full power days of operation at 5 MW, which would be drastically reduced if the reactor power is increased to 10 MW, as foreseen. We also show that under the proposed conditions, the experiment would not represent a significant penalty to the reactor normal operation. (author)

Determining prestressing forces for inspection of prestressed concrete containments

International Nuclear Information System (INIS)

1990-07-01

General Design Criterion 53, ''Provisions for Containment Testing and Inspection,'' of Appendix A, ''General Design Criteria for Nuclear Power Plants,'' to 10 CFR Part 50, ''Domestic Licensing of Production and Utilization Facilities,'' requires, in part, that the reactor containment be designed to permit (1) periodic inspection of all important areas and (2) an appropriate surveillance program. Regulatory Guide 1.35, ''Inservice Inspection of Ungrouted Tendons in Prestressed Concrete Containment Structures,'' describes a basis acceptable to the NRC staff for developing an appropriate inservice inspection and surveillance program for ungrouted tendons in prestressed concrete containment structures of light-water-cooled reactors. This guide expands and clarifies the NRC staff position on determining prestressing forces to be used for inservice inspections of prestressed concrete containment structures

Hermetic cable penetrations for containments of nuclear power reactors meet high safety standards

International Nuclear Information System (INIS)

Kusserow, J.; Gurr, W.; Pflug, H.

1985-05-01

Different types of cable penetrations for containments of nuclear power reactors have been developed and fabricated in the GDR. The technical parameters achieved are in accordance with the radiation protection requirements

Reactor

International Nuclear Information System (INIS)

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

1979-01-01

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

A user's guide to the reactor containment module NECTAR-SIRKIT

International Nuclear Information System (INIS)

Barker, C.D.

1982-02-01

The NECTAR environmental computer code has been developed to meet the increasing demand for comprehensive calculations of the radiological consequences of atmospheric releases of radioactivity. The code contains five calculational modules and this report presents a user's guide to the reactor containment module NECTAR-SIRKIT. The mathematical models employed in NECTAR-SIRKIT are briefly described and a complete specification of the input data required by the module is given. These data include facilities for specifying the fuel inventory, the release from irradiated fuel, primary and secondary containment parameters, filtration parameters and commands to identify the lineprinter output required. Three sample test cases are included in an Appendix to demonstrate some of the different ways in which the program may be used and also to provide examples for the prospective user. (author)

Leak detecting and identifying device in a reactor container

International Nuclear Information System (INIS)

Ito, Toshiichiro; Tomisawa, Teruaki; Yamada, Minoru.

1987-01-01

Purpose: To facilitate early detection and position identification for the leakages in a reactor container, shorten the start-up time for the nuclear power plant and reduce the equipment damages due to leakage. Constitution: Sensor signals from image sensors for obtaining infrared radiation image data are converted into image information and sent to a diagnosis device. While on the other hand, process variant signals from a process computer for obtaining plant status data are sent to a status judging device by which the plant status is judged based on the process variants such as water level, pressure and radioactivity in the reactor. The status judging device retrieves the status image aligned with the present plant status sent from the first memory device and transfers reference image information signals to the diagnosis device as the reference. The diagnosis device compares the present images with the reference images and displays the result of the judgement on CRT. (Yoshino, Y.)

Fiber reinforced concrete as a material for nuclear reactor containment buildings

International Nuclear Information System (INIS)

Mallikarjuna; Banthia, N.; Mindess, S.

1991-01-01

The fiber reinforced concrete as a constructional material for nuclear reactor containment buildings calls for an examination of its individual characteristics and potentialities due to its inherent superiority over normal plain and reinforced concrete. In the present investigation, first, to study the static behavior of straight, hooked-end and crimped fibers, recently developed nonlinear three-dimensional interface (contact) element has been used in conjunction with the eight nodded hexahedron and two nodded bar elements for concrete and steel fiber respectively. Then impact tests were carried out on fiber reinforced concrete beams with an instrumented drop weight impact machine. Two different concrete mixes were tested: normal strength and high strength concrete specimens. Fibers in the concrete mix found to significantly increase the ductility and the impact resistance of the composite. Deformed fibers increase peak pull-out load and pull-out distance, and perform better in the steel fiber reinforced concrete (SFRC) structures. (author)

Reactor operation safety information document

Energy Technology Data Exchange (ETDEWEB)

1990-01-01

The report contains a reactor facility description which includes K, P, and L reactor sites, structures, operating systems, engineered safety systems, support systems, and process and effluent monitoring systems; an accident analysis section which includes cooling system anomalies, radioactive materials releases, and anticipated transients without scram; a summary of onsite doses from design basis accidents; severe accident analysis (reactor core disruption); a description of operating contractor organization and emergency planning; and a summary of reactor safety evolution. (MB)

Modification method for increasing inner capacity of PWR type reactor container

International Nuclear Information System (INIS)

Toda, Shiro; Matsumoto, Mitsutaka

1998-01-01

A reactor container has a cylindrical wall portion and a dome-like ceiling portion. The upper portion of the cylindrical wall portion where the cross section is gradually reduced is cut off to from a circular step portion. A steel plate having an arcuate cross section having an inner surface being in contact with the outer circumferential surface of the ceiling portion is placed on the annular step portion, joined by welding to form a raised cylindrical portion. A raised ceiling portion having the same shape as the ceiling portion is constituted by weld-assembling of steel plates on the upper portion of the raised cylindrical portion. Through holes for air ventilation are perforated on the ceiling portion, and concrete is placed on the outer sides of the cylindrical wall portion, the raised cylindrical portion and the raised ceiling portion to form a concrete layer. Since the raised cylindrical portion and the raised ceiling portion can thus be constituted with no access to the inside of an existent reactor container, radiation exposure can be avoided. (I.N.)

The determination of load-time curve for a reactor containment which undergoes the impact of an airplane

International Nuclear Information System (INIS)

Wang Yuangong; Yu Aiping

1991-01-01

It is apparent that there exist some differences between the rigid load function of a reactor containment and the verified load function for undisturbed zone of deformative target when they undergo the impact of an airplane. The local non-linear behavior of impact zone is considered in the later situation, and a modified and better load function is obtained, which gives a value of the dynamic response of the structure 30% to 50% lower. In order to make engineering design more convenient and to simplify the calculation of non-linear behavior of the local zone, and based on the verified load function, several polygonal approximate curves are suggested for calculating the load functions of some typical structures

Treatment of opium alkaloid containing wastewater in sequencing batch reactor (SBR)-Effect of gamma irradiation

International Nuclear Information System (INIS)

Bural, Cavit B.; Demirer, Goksel N.; Kantoglu, Omer; Dilek, Filiz B.

2010-01-01

Aerobic biological treatment of opium alkaloid containing wastewater as well as the effect of gamma irradiation as pre-treatment was investigated. Biodegradability of raw wastewater was assessed in aerobic batch reactors and was found highly biodegradable (83-90% degradation). The effect of irradiation (40 and 140 kGy) on biodegradability was also evaluated in terms of BOD 5 /COD values and results revealed that irradiation imparted no further enhancement in the biodegradability. Despite the highly biodegradable nature of wastewater, further experiments in sequencing batch reactors (SBR) revealed that the treatment operation was not possible due to sludge settleability problem observed beyond an influent COD value of 2000 mg dm -3 . Possible reasons for this problem were investigated, and the high molecular weight, large size and aromatic structure of the organic pollutants present in wastewater was thought to contribute to poor settleability. Initial efforts to solve this problem by modifying the operational conditions, such as SRT reduction, failed. However, further operational modifications including addition of phosphate buffer cured the settleability problem and influent COD was increased up to 5000 mg dm -3 . Significant COD removal efficiencies (>70%) were obtained in both SBRs fed with original and irradiated wastewaters (by 40 kGy). However, pre-irradiated wastewater provided complete thebain removal and a better settling sludge, which was thought due to degradation of complex structure by radiation application. Degradation of the structure was observed by GC/MS analyses and enhancement in filterability tests.

Treatment of opium alkaloid containing wastewater in sequencing batch reactor (SBR)-Effect of gamma irradiation

Energy Technology Data Exchange (ETDEWEB)

Bural, Cavit B.; Demirer, Goksel N. [Middle East Technical University, Department of Environmental Engineering, 06531 Ankara (Turkey); Kantoglu, Omer [Turkish Atomic Energy Authority, Saraykoy Nuclear Research and Training Center, 06982, Kazan, Ankara (Turkey); Dilek, Filiz B., E-mail: fdilek@metu.edu.t [Middle East Technical University, Department of Environmental Engineering, 06531 Ankara (Turkey)

2010-04-15

Aerobic biological treatment of opium alkaloid containing wastewater as well as the effect of gamma irradiation as pre-treatment was investigated. Biodegradability of raw wastewater was assessed in aerobic batch reactors and was found highly biodegradable (83-90% degradation). The effect of irradiation (40 and 140 kGy) on biodegradability was also evaluated in terms of BOD{sub 5}/COD values and results revealed that irradiation imparted no further enhancement in the biodegradability. Despite the highly biodegradable nature of wastewater, further experiments in sequencing batch reactors (SBR) revealed that the treatment operation was not possible due to sludge settleability problem observed beyond an influent COD value of 2000 mg dm{sup -3}. Possible reasons for this problem were investigated, and the high molecular weight, large size and aromatic structure of the organic pollutants present in wastewater was thought to contribute to poor settleability. Initial efforts to solve this problem by modifying the operational conditions, such as SRT reduction, failed. However, further operational modifications including addition of phosphate buffer cured the settleability problem and influent COD was increased up to 5000 mg dm{sup -3}. Significant COD removal efficiencies (>70%) were obtained in both SBRs fed with original and irradiated wastewaters (by 40 kGy). However, pre-irradiated wastewater provided complete thebain removal and a better settling sludge, which was thought due to degradation of complex structure by radiation application. Degradation of the structure was observed by GC/MS analyses and enhancement in filterability tests.

Non-intuitive fluid dynamics from reactor and containment technology

International Nuclear Information System (INIS)

Moody, F.J.

1986-01-01

One exciting aspect of fluid dynamics is that the subject has many surprises. The surprises can be good, but if not anticipated, they sometimes can be costly and embarrassing. Several non-intuitive fluid responses have emerged from studies in nuclear reactor and containment design. These responses include bubble behavior, blowdown, and waterhammer phenomena. Apologies are extended to those who are not surprised by the results. However, many will find the examples interesting; some have been amazed; a few have declared a personal crisis in their engineering perception

Development of a coupled containment-reactor coolant system methodology for the analysis of IRIS small break LOCA

International Nuclear Information System (INIS)

Manfredini, Antonio; Oriolo, Francesco; Paci, Sandro; Oriani, Luca

2003-01-01

The main purpose of the present work is to identify the most relevant physical phenomena for the IRIS (International Reactor Innovative and Secure) containment system and the development of an integrated methodology for the simultaneous safety analysis of both the reactor and containment with available computer codes. Specific objectives are: (a) to assess the limitations of the lumped parameter codes on predictions of complex situations; (b) to identify alternatives to classical containment analysis techniques. The characteristic features of an integral reactor like IRIS present a much greater challenge to code developers than conventional, loop type PWRs. In particular, the integral primary system and the containment are strongly coupled during postulated accident conditions and thus an integrated simulation of both systems is required to obtain a reliable phenomenological representation. The comparison of the results obtained in the application of two containment codes (GOTHIC and integrated FUMO) on 'ad hoc' IRIS related benchmarks will also be described. These preliminary calculations were used to test the IRIS containment concept and cooling strategies, at the same time highlighting the most relevant issues that require a more refined investigation. Finally, this activity allowed to perform more refined calculations, in progress at the moment, aimed at showing that the IRIS safety systems and containment design solutions perform their intended functions. (author)

Modeling of thermophoretic deposition of aerosols in nuclear reactor containments

International Nuclear Information System (INIS)

Fernandes, A.; Loyalka, S.K.

1996-01-01

Aerosol released in postulated or real nuclear reactor accidents can deposit on containment surfaces via motion induced by temperature gradients in addition to the motion due to diffusion and gravity. The deposition due to temperature gradients is known as thermophoretic deposition, and it is currently modeled in codes such as CONTAIN in direct analogy with heat transfer, but there have been questions about such analogies. This paper focuses on a numerical solution of the particle continuity equation in laminar flow condition characteristics of natural convection. First, the thermophoretic deposition rate is calculated as a function of the Prandtl and Schmidt numbers, the thermophoretic coefficient K, and the temperature difference between the atmosphere and the wall. Then, the cases of diffusion alone and a boundary-layer approximation (due to Batchelor and Shen) to the full continuity equation are considered. It is noted that an analogy with heat transfer does not hold, but for the circumstances considered in this paper, the deposition rates from the diffusion solution and the boundary-layer approximation can be added to provide reasonably good agreement (maximum deviation 30%) with the full solution of the particle continuity equation. Finally, correlations useful for implementation in the reactor source term codes are provided

Hydrogen management in nuclear reactor containment

International Nuclear Information System (INIS)

Iyer, Kannan

2014-01-01

The talk will present the systematic methodology evolved to assess the hydrogen management in nuclear reactor containment during a severe accident. The focus is on the methodology evolved as the full problem is yet to be solved completely. First, the method to quantify mixing of hydrogen is presented. It is demonstrated that buoyancy modified model is adequate to quantify the process satisfactorily. On noting that the hydrogen levels are higher than the safe limits, effort was directed towards mitigating the concentration. Passive Auto-catalytic Recombiners (PAR) were identified as the potential devices for mitigation. Efforts were then directed to model these and a satisfactory one-step reaction derived from a 12 reaction model was evolved. This model was satisfactory when compared with experimental results with hydrogen concentration below 4%. However, the same when extended to hydrogen concentration of 20%, predicts very high concentration thereby indicating the need for experiments at high hydrogen concentration. (author)

Nuclear reactor apparatus

International Nuclear Information System (INIS)

Braun, H.E.; Bonnet, H.P.

1978-01-01

The reactor and its containment, instead of being supported on a solid concrete pad, are supported on a truss formed of upper and lower reinforced horizontal plates and vertical walls integrated into a rigid structure. The plates and walls from chambers within which the auxiliary components of the reactor, such as valves, pumping equipment and various tanks, are disposed. Certain of the chambers are also access passages for personnel, pipe chases, valve chambers and the like. In particular the truss includes an annular chamber. This chamber is lined and sealed by a corrosion-resistant liner and contains coolant and serves as a refueling cooling storage tank. This tank is directly below the primary-coolant conductor loops which extend from the reactor above the upper plate. The upper plate includes a sump connected to the tank through which coolant flows into the tank in the event of the occurrence of a loss-of-coolant accident. The truss extends beyond the containment and has chambers in the extending annulus. Pumps for circulating the coolant between the refueling coolant storage tank and the reactor are provided in certain of these chambers. The pumps are connected to the reactor by relatively short coolant conductors. Access to these pumps is readily afforded through hatches in the extending annulus

Validation of containment thermal hydraulic computer codes for VVER reactor

Energy Technology Data Exchange (ETDEWEB)

Jiri Macek; Lubomir Denk [Nuclear Research Institute Rez plc Thermal-Hydraulic Analyses Department CZ 250 68 Husinec-Rez (Czech Republic)

2005-07-01

Full text of publication follows: The Czech Republic operates 4 VVER-440 units, two VVER-1000 units are being finalized (one of them is undergoing commissioning). Thermal-hydraulics Department of the Nuclear Research Institute Rez performs accident analyses for these plants using a number of computer codes. To model the primary and secondary circuits behaviour the system codes ATHLET, CATHARE, RELAP, TRAC are applied. Containment and pressure-suppression system are modelled with COCOSYS and MELCOR codes, the reactor power calculations (point and space-neutron kinetics) are made with DYN3D, NESTLE and CDF codes (FLUENT, TRIO) are used for some specific problems.An integral part of the current Czech project 'New Energy Sources' is selection of a new nuclear source. Within this and the preceding projects financed by the Czech Ministry of Industry and Trade and the EU PHARE, the Department carries and has carried out the systematic validation of thermal-hydraulic and reactor physics computer codes applying data obtained on several experimental facilities as well as the real operational data. One of the important components of the VVER 440/213 NPP is its containment with pressure suppression system (bubble condenser). For safety analyses of this system, computer codes of the type MELCOR and COCOSYS are used in the Czech Republic. These codes were developed for containments of classic PWRs or BWRs. In order to apply these codes for VVER 440 systems, their validation on experimental facilities must be performed.The paper provides concise information on these activities of the NRI and its Thermal-hydraulics Department. The containment system of the VVER 440/213, its functions and approaches to solution of its safety is described with definition of acceptance criteria. A detailed example of the containment code validation on EREC Test facility (LOCA and MSLB) and the consequent utilisation of the results for a real NPP purposes is included. An approach to

Investigation of structural materials of reactors using high-energy heavy-ion irradiations

International Nuclear Information System (INIS)

Wang Zhiguang

2007-01-01

Radiation damage in structural materials of fission/fusion reactors is mainly attributed to the evolution of intensive atom displacement damage induced by energetic particles (n, α and/or fission fragments) and high-rate helium doping by direct α particle bombardments and/or (n, α) reactions. It can cause severe degradation of reactor structural materials such as surface blistering, bulk void swelling, deformation, fatigue, embrittlement, stress erosion corrosion and so on that will significantly affect the operation safety of reactors. However, up to now, behavior of structural materials at the end of their service can hardly be fully tested in a real reactor. In the present work, damage process in reactor structural materials is briefly introduced, then the advantages of energetic ion implantation/irradiation especially high-energy heavy ion irradiation are discussed, and several typical examples on simulation of radiation effects in reactor candidate structural materials using high-energy heavy ion irradiations are pronounced. Experimental results and theoretical analysis suggested that irradiation with energetic particles especially high-energy heavy ions is very useful technique for simulating the evolution of microstructures and macro-properties of reactor structural materials. Furthermore, an on-going plan of material irradiation experiments using high energy H- and He-ions based on the Heavy Ion Research Facilities in Lanzhou (HIRFL) is also briefly interpreted. (authors)

Containment safety margins

International Nuclear Information System (INIS)

Von Riesemann, W.A.

1980-01-01

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

Fabrication and Pre-irradiation Characterization of a Minor Actinide and Rare Earth Containing Fast Reactor Fuel Experiment for Irradiation in the Advanced Test Reactor

Energy Technology Data Exchange (ETDEWEB)

Timothy A. Hyde

2012-06-01

The United States Department of Energy, seeks to develop and demonstrate the technologies needed to transmute the long-lived transuranic actinide isotopes contained in spent nuclear fuel into shorter lived fission products, thereby decreasing the volume of material requiring disposal and reducing the long-term radiotoxicity and heat load of high-level waste sent to a geologic repository. This transmutation of the long lived actinides plutonium, neptunium, americium and curium can be accomplished by first separating them from spent Light Water Reactor fuel using a pyro-metalurgical process, then reprocessing them into new fuel with fresh uranium additions, and then transmuted to short lived nuclides in a liquid metal cooled fast reactor. An important component of the technology is developing actinide-bearing fuel forms containing plutonium, neptunium, americium and curium isotopes that meet the stringent requirements of reactor fuels and materials.

Prospect of Ti-Ni shape memory alloy applied in reactor structures

International Nuclear Information System (INIS)

Duan Yuangang

1995-01-01

Shape memory effect mechanism, physical property, composition, manufacturing process and application in mechanical structure of Ti-Ni shape memory alloy are introduced. Applications of Ti-Ni shape memory alloy in reactor structure are prospected and some necessary technical conditions of shape memory alloy applied in the reactor structure are put forward initially

An Evaluation Report on the High Temperature Design of the KALIMER-600 Reactor Structures

Energy Technology Data Exchange (ETDEWEB)

Park, Chang Gyu; Lee, Jae Han

2007-03-15

This report is on the validity evaluation of high temperature structural design for the reactor structures and piping of the pool-type Liquid Metal Reactor, KALIMER-600 subjected to the high temperature thermal load condition. The structural concept of the Upper Internal Structure located above the core is analyzed and the adequate UIS conceptual design for KALIMER-600 is proposed. Also, the high temperature structural integrity of the thermal liner which is to protect the UIS bottom plate from the high frequency thermal fatigue damage was evaluated by the thermal stripping analysis. The high temperature structural design of the reactor internal structure by considering the reactor startup-shutdown cycle was carried out and the structural integrity of it for a normal operating condition as well as the transient condition of the primary pump trip accident was confirmed. Additionally the structure design of the reactor internal structural was changed to prevent the non-uniform deformation of the primary pump which is induced by the thermal expansion difference between the reactor head and the baffle plate. The arrangement of the IHTS piping system which is a part of the reactor system is carried out and the structural integrity and the accumulated deformation by considering the reactor startup-shutdown cycle of a normal operating condition were evaluated. The structural integrity and the accumulated deformation of the PDRC hot leg piping by considering the PDRC operating condition were evaluated. The validity of KALIMER-600 high temperature structural design is confirmed through this study, and it is clearly found that the methodology research to evaluate the structural integrity considering the reactor life time of 60 years ensured is necessary.

Experimental and theoretical investigations of soil-structure interaction effect at HDR-reactor-building

International Nuclear Information System (INIS)

Wassermann, K.

1983-01-01

Full-scale dynamic testing on intermediate and high levels was performed at the Heissdampfreaktor (HDR) in 1979. Various types of dynamic forces were applied and response of the reactor containment structure and internal components was measured. Precalculations of dynamic behaviour and response of the structure were made through different mathematical models for the structure and the underlying soil. Soil-Structure Interaction effects are investigated and different theoretical models are compared with experimental results. In each model, the soil is represented by springs attached to the structural model. Stiffnesses of springs are calculated by different finite-element idealizations and half-space approximations. Eigenfrequencies and damping values related to interaction effects (translation, rocking, torsion) are identified from test results. The comparisons of dynamic characteristic of the soil-structure system between precalculations and test results show good agreement in general. (orig.)

Coupled fluid/structure response of a reactor cover to slug impact loading

International Nuclear Information System (INIS)

Smith, B.L.; Saurer, G.; Wanner, R.; Palsson, H.

1983-05-01

The response of an LMFBR roof structure to slug impact loads is investigated using a combined 2D and 3D approach based on the containment code SEURBNUK and the finite element structure code ADINA. A specimen roof design of box-type construction with concrete infill is adopted for the study, with dimensions appropriate to a commercial-sized fast reactor of the 'pool' type. Provision is made in the model for the location of the major roof penetrations, and the roof annulus is closed in the central section by a rigid, but movable plug concentric with the axis of symmetry. An interface between the codes SEURBNUK and ADINA is made possible by defining a 2D substitute roof model with material properties chosen to match the principal response characteristics of the detailed model. The SEURBNUK code, recently extended to account for coupling of roof loading and roof response, uses the 2D model, incorporated in an appropriate reactor geometry, to examine the fluid-structure interactions and to supply roof pressure loadings for the ADINA runs. A strategy for cross-checking the structural equivalence of the 2D and 3D roof models is developed, and this operates in parallel with the loading and response computations. The first exploratory SEURBNUK calculations are described in which the roof is represented by a simple homogeneous plate. (Auth.)

RB Research nuclear reactor RB reactor, Annual report for 2000

International Nuclear Information System (INIS)

Milosevic, M.

2000-12-01

Report on RB reactor operation during 2000 contains 3 parts. Part one contains a brief description of reactor operation and reactor components, relevant dosimetry data and radiation protection issues, personnel and financial data. Part two is devoted to maintenance of the reactor components, namely, fuel, heavy water, reactor vessel, heavy water circulation system, absorption rods and heavy water level-meters, maintenance of electronic, mechanical, electrical and auxiliary equipment. Part three contains data concerned with reactor operation and utilization with a comprehensive list of publications resulting from experiments done at the RB reactor. It contains data about reactor operation during previous 14 years, i.e. from 1986 - 2000

Status of advanced containment systems for next generation water reactors

International Nuclear Information System (INIS)

1994-06-01

The present IAEA status report is intended to provide information on the current status and development of containment systems of the next generation reactors for electricity production and, particularly, to highlight features which may be considered advanced, i.e. which present improved performance with evolutionary or innovative design solutions or new design approaches. The objectives of the present status report are: To present, on a concise and consistent basis, selected containment designs currently being developed in the world; to review and compare new approaches to the design bases for the containments, in order to identify common trends, that may eventually lead to greater worldwide consensus, to identify, list and compare existing design objectives for advanced containments, related to safety, availability, maintainability, plant life, decommissioning, economics, etc.; to describe the general approaches adopted in different advanced containments to cope with various identified challenges, both those included in the current design bases and those related to new events considered in the design; to briefly identify recent achievements and future needs for new or improved computer codes, standards, experimental research, prototype testing, etc. related to containment systems; to describe the outstanding features of some containments or specific solutions proposed by different parties and which are generally interesting to the international scientific community. 36 refs, 27 figs, 1 tab

Shield structure for a nuclear reactor

International Nuclear Information System (INIS)

Rouse, C.A.; Simnad, M.T.

1979-01-01

An improved nuclear reactor shield structure is described for use where there are significant amounts of fast neutron flux above an energy level of approximately 70 keV. The shield includes structural supports and neutron moderator and absorber systems. A portion at least of the neutron moderator material is magnesium oxide either alone or in combination with other moderator materials such as graphite and iron. (U.K.)

Structural and piping issues in the design certification of advanced reactors

International Nuclear Information System (INIS)

Ali, S.A.; Terao, D.; Bagchi, G.

1996-01-01

The purpose of this paper is to discuss the design certification of structures and piping for evolutionary and passive advanced light water reactors. Advanced reactor designs are based on a set of assumed site-related parameters that are selected to envelop a majority of potential nuclear power plant sites. Multiple time histories are used as the seismic design basis in order to cover the majority of potential sites in the US. Additionally, design are established to ensure that surface motions at a particular site will not exceed the enveloped standard design surface motions. State-of-the-art soil-structure interaction (SSI) analyses have been performed for the advanced reactors, which include structure-to-structure interaction for all seismic Category 1 structures. Advanced technology has been utilized to exclude the dynamic effects of pipe rupture from structural design by demonstrating that the probability of pipe rupture is extremely low. For piping design, the advanced reactor vendors have developed design acceptance criteria (DAC) which provides the piping design analysis methods, design procedures, and acceptance criteria. In SECY-93-087 the NRC staff recommended that the Commission approve the approach to eliminate the OBE from the design of structures and piping in advanced reactors and provided guidance which identifies the necessary changes to existing seismic design criteria. The supplemental criteria address fatigue, seismic anchor motion, and piping stress limits when the OBE is eliminated

Testing of a steel containment vessel model

International Nuclear Information System (INIS)

Luk, V.K.; Hessheimer, M.F.; Matsumoto, T.; Komine, K.; Costello, J.F.

1997-01-01

A mixed-scale containment vessel model, with 1:10 in containment geometry and 1:4 in shell thickness, was fabricated to represent an improved, boiling water reactor (BWR) Mark II containment vessel. A contact structure, installed over the model and separated at a nominally uniform distance from it, provided a simplified representation of a reactor shield building in the actual plant. This paper describes the pretest preparations and the conduct of the high pressure test of the model performed on December 11-12, 1996. 4 refs., 2 figs

Treatment of opium alkaloid containing wastewater in sequencing batch reactor (SBR)—Effect of gamma irradiation

Science.gov (United States)

Bural, Cavit B.; Demirer, Goksel N.; Kantoglu, Omer; Dilek, Filiz B.

2010-04-01

Aerobic biological treatment of opium alkaloid containing wastewater as well as the effect of gamma irradiation as pre-treatment was investigated. Biodegradability of raw wastewater was assessed in aerobic batch reactors and was found highly biodegradable (83-90% degradation). The effect of irradiation (40 and 140 kGy) on biodegradability was also evaluated in terms of BOD 5/COD values and results revealed that irradiation imparted no further enhancement in the biodegradability. Despite the highly biodegradable nature of wastewater, further experiments in sequencing batch reactors (SBR) revealed that the treatment operation was not possible due to sludge settleability problem observed beyond an influent COD value of 2000 mg dm -3. Possible reasons for this problem were investigated, and the high molecular weight, large size and aromatic structure of the organic pollutants present in wastewater was thought to contribute to poor settleability. Initial efforts to solve this problem by modifying the operational conditions, such as SRT reduction, failed. However, further operational modifications including addition of phosphate buffer cured the settleability problem and influent COD was increased up to 5000 mg dm -3. Significant COD removal efficiencies (>70%) were obtained in both SBRs fed with original and irradiated wastewaters (by 40 kGy). However, pre-irradiated wastewater provided complete thebain removal and a better settling sludge, which was thought due to degradation of complex structure by radiation application. Degradation of the structure was observed by GC/MS analyses and enhancement in filterability tests.

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)

Structural analysis of the Upper Internals Structure for the Clinch River Breeder Reactor Plant

International Nuclear Information System (INIS)

Houtman, J.L.

1979-01-01

The Upper Internals Structure (UIS) of the Clinch River Breeder Reactor Plant (CRBRP) provides control of core outlet flow to prevent severe thermal transients from occuring at the reactor vessel and primary heat transport outlet piping, provides instrumentation to monitor core performance, provides support for the control rod drivelines, and provides secondary holddown of the core. All of the structural analysis aspects of assuring the UIS is structurally adequate are presented including simplified and rigorous inelastic analysis methods, elevated temperature criteria, environmental effects on material properties, design techniques, and manufacturing constraints

Fuel, structural material and coolant for an advanced fast micro-reactor

International Nuclear Information System (INIS)

Nascimento, Jamil A. do; Guimaraes, Lamartine N.F.; Ono, Shizuca

2011-01-01

The use of nuclear reactors in space, seabed or other Earth hostile environment in the future is a vision that some Brazilian nuclear researchers share. Currently, the USA, a leader in space exploration, has as long-term objectives the establishment of a permanent Moon base and to launch a manned mission to Mars. A nuclear micro-reactor is the power source chosen to provide energy for life support, electricity for systems, in these missions. A strategy to develop an advanced micro-reactor technologies may consider the current fast reactor technologies as back-up and the development of advanced fuel, structural and coolant materials. The next generation reactors (GEN-IV) for terrestrial applications will operate with high output temperature to allow advanced conversion cycle, such as Brayton, and hydrogen production, among others. The development of an advanced fast micro-reactor may create a synergy between the GEN-IV and space reactor technologies. Considering a set of basic requirements and materials properties this paper discusses the choice of advanced fuel, structural and coolant materials for a fast micro-reactor. The chosen candidate materials are: nitride, oxide as back-up, for fuel, lead, tin and gallium for coolant, ferritic MA-ODS and Mo alloys for core structures. The next step will be the neutronic and burnup evaluation of core concepts with this set of materials. (author)

Treatment of EDTA contained reactor coolant using water dielectric barrier discharge plasma

International Nuclear Information System (INIS)

Song, Sang Heon; Kwon, Daniel; Kim, Gon Ho

2005-01-01

EDTA (Ethylene Diamine Tetraacetic Acid) is used as a main absorbent for the metal ion in the secondary loop of the nuclear reactor. Dissolving the wasted EDTA with low cost, therefore, is important issue for the maintenance of the nuclear power reactor and the protection of environment. EDTA is not easily biodegradable, furthermore these methods could make remained another pollutant as complex chemical compounds. Compared to chemical method, the physical methods, using the energetic particles and UVs, are more favorable because they dissociate the bonds of organic compounds directly without the secondary chemical reactions during the treatment. Recently, high energy electron beam, the plasma torch, and the water breakdown by high voltage pulse are applied to treatment of the waste water contained chemicals. Here consideration is narrow down to improve the interaction between the plasma and the chemical bonds of EDTA because the energetic particles; activated radicals, and UVs, are abundant in plasmas. The new method adapted of the water DBD (dielectric barrier discharge) which plasma generates directly on the top of the water contained EDTA is proposed. The application of DBD plasmas has been extended for cleaning the organic compounds from the contaminated surface and also for removing volatile organic chemicals (VOC) such as NO x and SO x from the exhausted gases. Here, the water DBD reactor (SEMTECH, SD-DWG-04-1) is consisted that the one electrode is a ceramic insulator and another one is the water itself. Interestingly, the one electrode, the water, is not the solid dielectric electrode. In this study, therefore, the characteristics with driving frequency are considered and the feasibility of this new method for the DBD treatment of EDTA contained water is demonstrated

Preliminary structural evaluations of the STAR-LM reactor vessel and the support design

International Nuclear Information System (INIS)

Koo, Gyeong-Hoi; Sienicki, James J.; Moisseytsev, Anton

2007-01-01

In this paper, preliminary structural evaluations of the reactor vessel and support design of the STAR-LM (The Secure, Transportable, Autonomous Reactor - Liquid Metal variant), which is a lead-cooled reactor, are carried out with respect to an elevated temperature design and seismic design. For an elevated temperature design, the structural integrity of a direct coolant contact to the reactor vessel is investigated by using a detail structural analysis and the ASME-NH code rules. From the results of the structural analyses and the integrity evaluations, it was found that the design concept of a direct coolant contact to the reactor vessel cannot satisfy the ASME-NH rules for a given design condition. Therefore, a design modification with regards to the thermal barrier is introduced in the STAR-LM design. For a seismic design, detailed seismic time history response analyses for a reactor vessel with a consideration of a fluid-structure interaction are carried out for both a top support type and a bottom support type. And from the results of the hydrodynamic pressure responses, an investigation of the minimum thickness design of the reactor vessel is tentatively carried out by using the ASME design rules

The economic potential of a cassette-type-reactor-installed nuclear ice-breaking container ship

International Nuclear Information System (INIS)

Kondo, Koichi; Takamasa, Tomoji

1999-01-01

An improved cassette-type marine reactor MRX (Marine Reactor X) which is currently researched and developed by the Japan Atomic Energy Research Institute is designed to be easily removed and transferred to another ship. If the reactor in a nuclear-powered ship, which is the reason for its higher cost, were replaced by the cassette-type-MRX, the reusability of the MRX would reduce the cost difference between nuclear-powered and diesel ships. As an investigation of one aspect of a cassette-type MRX, we attempted in this study to do an economic review of an MRX-installed nuclear-powered ice-breaking container ship sailing via the Arctic Ocean. The transportation cost between the Far East and Europe to carry one TEU (twenty-foot-equivalent container unit) over the entire life of the ship for an MRX (which is used for a 20-year period)-installed container ship sailing via the Arctic Ocean is about 70% higher than the Suez Canal diesel ship, carrying 8,000 TEU and sailing at 25 knots, and about 10% higher than the Suez Canal diesel ship carrying 4,000 TEU and sailing at 34 knots. The cost for a cassette-type-MRX (which is used for a 40-year period, removed and transferred to a second ship after being used for 20 years in the first ship)-installed nuclear-powered container ship is about 7% lower than that for the one operated for 20 years. Considering any loss or reduction in sales opportunities through the extension of the transportation period, the nuclear-powered container ship via the Arctic Sea is a more suitable means of transportation than a diesel ship sailing at 25 knots via the Suez Canal when the value of the commodities carried exceeds 2,800 dollars per freight ton. (author)

Nuclear reactor having an inflatable vessel closure seal structure

International Nuclear Information System (INIS)

1980-01-01

An improved type of closure head seal for the rotatable plugs of the reactor vessel of a liquid metal fast breeder reactor is described. The seal prevents the release of radioactive particles while allowing the plug to be rotated without major manipulation of the seal structure. (UK)

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

Lining up device for the internal structures of a nuclear reactor vessel

International Nuclear Information System (INIS)

Silverblatt, B.L.

1977-01-01

The invention concerns a nuclear reactor of the type with a vessel, a vessel head carried at the top of this vessel by a core cylinder comprising a flange internally supported by the vessel, and an upper support structure supported between the core cylinder flange and the vessel head to align laterally the head, vessel, flange and support structure. A bottom key device is provided for lining up the flange, support structure and vessel, and an upper key device for laterally lining up support structure and the vessel head and for maintaining this alignment when they are removed simultaneously from the core cylinder and vessel. When re-assembling the reactor, the top support structure and the vessel head are lowered simultaneously so that an opening in the top alignment structure engages in the upper extension of the bottom alignment structure. A plurality of alignment stuctures may be utilised round the circumference of the reactor vessel. The disposition of the invention also facilitates the removal of the core cylinder from the reactor vessel. In this way, the alignment on re-assembly is ensured by the re-entry of the bottom extension under the flange of the core cylinder with the groove or keyway of the reactor vessel [fr

The surrounding concrete structure of the containment as a safety component

International Nuclear Information System (INIS)

Alex, H.; Kuntze, W.M.

1978-01-01

This paper will briefly discuss the containments of the various types of reactors in the Federal Republic of Germany and will try to show the importance of the surrounding concrete structures with respect to safety. It will be seen that the surrounding concrete structures serve in any case - as protection against external events - as secondary shielding and must therefore be considered as a passive safety feature. The design requirements for the surrounding concrete structures with respect to protection against external events and to physical protection generally supplement each other. Reference will be made to possible alternatives, which might result from studies of underground siting of nuclear power plants. Whether or not this type of construction can lead to additional safety can only be judged when the results of all these studies - some of which are still under way - are evaluated. The concluding part of this paper will deal with the responsibilities of the civil engineering supervisory authorities and the nuclear licensing authorities with respect to the surrounding concrete structures. (orig.) [de

A study on ex-vessel steam explosion for a flooded reactor cavity of reactor scale - 15216

International Nuclear Information System (INIS)

Song, S.; Yoon, E.; Kim, Y.; Cho, Y.

2015-01-01

A steam explosion can occur when a molten corium is mixed with a coolant, more volatile liquid. In severe accidents, corium can come into contact with coolant either when it flows to the bottom of the reactor vessel and encounters the reactor coolant, or when it breaches the reactor vessel and flows into the reactor containment. A steam explosion could then threaten the containment structures, such as the reactor vessel or the concrete walls/penetrations of the containment building. This study is to understand the shortcomings of the existing analysis code (TEXAS-V) and to estimate the steam explosion loads on reactor scale and assess the effect of variables, then we compared results and physical phenomena. Sensitivity study of major parameters for initial condition is performed. Variables related to melt corium such as corium temperature, falling velocity and diameter of melt are more important to the ex-vessel steam explosion load and the steam explosion loads are proportional to these variables related to melt corium. Coolant temperature on reactor cavity has a specific area to increase the steam explosion loads. These results will be used to evaluate the steam explosion loads using ROAAM (Risk Oriented Accident Analysis Methodology) and to develop the evaluation methodology of ex-vessel steam explosion. (authors)

Analysis of soil-structure interaction and floor response spectrum of reactor building for China advanced research reactor

International Nuclear Information System (INIS)

Rong Feng; Wang Jiachun; He Shuyan

2006-01-01

Analysis of Soil-Structure Interaction (SSI) and calculation of Floor Response Spectrum (FRS) is substantial for anti-seismic design for China Advanced Research Reactor (CARR) project. The article uses direct method to analyze the seismic reaction of the reactor building in considering soil-structure interaction by establishing two-dimensional soil-structure co-acting model for analyzing and inputting of seismic waves from three directions respectively. The seismic response and floor response spectrum of foundation and floors of the building under different cases have been calculated. (authors)

Procedures of ASME code case N-201 for KALIMER. Reactor internal structures

International Nuclear Information System (INIS)

Koo, Gyeong Hoi; Yoo, B.

2001-02-01

The main objective of this report is to describe the design procedure of ASME Boiler and Pressure Vessel Code, Code Case N-201-4, which is an elevated temperature structural design code of the Nuclear reactor internal structures, checking the criteria of stress limit, accumulated inelastic strain and deformation, creep-fatigue damage, and buckling limit. As one of examples, the creep-fatigue damage evaluations are carried out for the KALIMER reactor internal structures of baffle annulus. This report is expected to be very useful in evaluating the structural integrity of the liquid metal reactor operating under an elevated temperature

High-temperature-structural design and research and development for reactor system components

International Nuclear Information System (INIS)

Matsumura, Makoto; Hada, Mikio

1985-01-01

The design of reactor system components requires high-temperature-structural design guide with the consideration of the creep effect of materials related to research and development on structural design. The high-temperature-structural design guideline for the fast prototype reactor MONJU has been developed under the active leadership by Power Reactor and Nuclear Fuel Development Corporation and Toshiba has actively participated to this work with responsibility on in-vessel components, performing research and development programs. This paper reports the current status of high-temperature-structural-design-oriented research and development programs and development of analytical system including stress-evaluation program. (author)

The French nuclear power plant reactor building containment contributions of prestressing and concrete performances in reliability improvements and cost savings

International Nuclear Information System (INIS)

Rouelle, P.; Roy, F.

1998-01-01

The Electricite de France's N4 CHOOZ B nuclear power plant, two units of the world's largest PWR model (1450 Mwe each), has earned the Electric Power International's 1997 Powerplant Award. This lead NPP for EDF's N4 series has been improved notably in terms of civil works. The presentation will focus on the Reactor Building's inner containment wall which is one of the main civil structures on a technical and safety point of view. In order to take into account the necessary evolution of the concrete technical specification such as compressive strength low creep and shrinkage, the HSC/HPC has been used on the last N4 Civaux 2 NPP. As a result of the use of this type of professional concrete, the containment withstands an higher internal pressure related to severe accident and ensures higher level of leak-tightness, thus improving the overall safety of the NPP. On that occasion, a new type of prestressing has been tested locally through 55 C 15 S tendons using a new C 1500 FE Jack. These updated civil works techniques shall allow EDF to ensure a Reactor Containment lifespan for more than 50 years. The gains in terms of reliability and cost saving of these improved techniques will be developed hereafter

Modular reactor head shielding system

International Nuclear Information System (INIS)

Jacobson, E. B.

1985-01-01

An improved modular reactor head shielding system is provided that includes a frame which is removably assembled on a reactor head such that no structural or mechanical alteration of the head is required. The shielding system also includes hanging assemblies to mount flexible shielding pads on trolleys which can be moved along the frame. The assemblies allow individual pivoting movement of the pads. The pivoting movement along with the movement allowed by the trolleys provides ease of access to any point on the reactor head. The assemblies also facilitate safe and efficient mounting of the pads directly to and from storage containers such that workers have additional shielding throughout virtually the entire installation and removal process. The flexible shielding pads are designed to interleave with one another when assembled around the reactor head for substantially improved containment of radiation leakage

System of large transport containers for waste from dismantling light water and gas-cooled nuclear reactors. Volume 1

International Nuclear Information System (INIS)

Price, M.S.T.; Lafontaine, I.

1985-01-01

The purpose of this volume is to introduce the main types of nuclear reactor in the European Community (EC), select reference plants for further study, estimate the waste streams from the reference reactors, survey the transport regulations and assess existing containers

A reactor core/containment status evaluation flowchart for determining protective actions in emergencies

International Nuclear Information System (INIS)

Glissman, M.A.

1988-01-01

In the event of an emergency at a power reactor station, there might not be adequate time or sufficient data to fully assess radiological implications and make protective action recommendations based on projected population exposures. Thus, decision-making guidance is needed that is based on readily available plant indicators, not just on time-consuming dose calculations. In the United States, this guidance must be compatible with the recommended by the Nuclear Regulatory Commission and the Environmental Protection Agency, and it must include predetermined, measurable, site-specific parameters for assessing conditions in the reactor core and containment. The preparation of this real time guidance calls for the selection of suitable parameters and the determination of the values for these parameters that will correspond to different levels of protective action. This process is illustrated in this paper by selecting parameters and determining appropriate values for constructing a Core/Containment Status Evaluation Flowchart for an example power plant

Aging management of light water reactor concrete containments

International Nuclear Information System (INIS)

Shah, V.N.; Hookhman, C.J.

1994-01-01

This paper evaluates aging of light water reactor concrete containments and identifies three degradation mechanisms that have potential to cause widespread aging damage after years of satisfactory experience: alkali-silica reaction, corrosion of reinforcing steel, and sulfate attack. The evaluation is based on a comprehensive review of the relevant technical literature. Low-alkali cement and slow-reacting aggregates selected according to ASTM requirements cause deleterious alkali-silica reactions. Low concentrations of chloride ions can initiate corrosion of the reinforcing steel if the hydroxyl ions are sufficiently reduced by carbonation, leaching, or magnesium sulfate attack. Magnesium sulfate attack on concrete can cause loss of strength and cementitious properties after long exposure. Techniques to detect and mitigate these long-term aging effects are discussed

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.

Packed bed reactor for degradation of simulated cyanide-containing wastewater

OpenAIRE

Kumar, Virender; Kumar, Vijay; Bhalla, Tek Chand

2014-01-01

The discharge of cyanide-containing effluents into the environment contaminates water bodies and soil. Effective methods of treatment which can detoxify cyanide are the need of the hour. The aim of the present study is to develop a bioreactor for complete degradation of cyanide using immobilized cells of Serratia marcescens RL2b. Alginate-entrapped cells of S. marcescens RL2b were used for complete degradation of cyanide in a packed bed reactor (PBR). Cells grown in minimal salt medium (pH 6....

Tritium handling, breeding, and containment in two conceptual fusion reactor designs: UWMAK-II and UWMAK-III

International Nuclear Information System (INIS)

Clemmer, R.G.; Larsen, E.M.

1976-01-01

Tritium is an essential component of near-term controlled thermonuclear reactor systems. Since tritium is not likely to be available on a large scale at a modest cost, fusion reactor designs must incorporate blanket systems which will be capable of breeding tritium. Because of the radiological activity and capability of assimilation into living tissues, tritium release to the environment must be strictly controlled. The University of Wisconsin has completed three conceptual designs of fusion reactors, UWMAK-I, UWMAK-II, and UWMAK-III. This report discusses the tritium systems for UWMAK-II, a reactor design with a helium cooled solid breeder blanket, and UWMAK-III, a reactor design with a high-temperature liquid breeder blanket. Tritium systems for fueling and recycling, breeding and recovery, and plant containment and control are discussed. (Auth.)

Analysis of denitrifier community in a bioaugmented sequencing batch reactor for the treatment of coking wastewater containing pyridine and quinoline

Energy Technology Data Exchange (ETDEWEB)

Bai, Yaohui; Xing, Rui; Wen, Donghui; Tang, Xiaoyan [Peking Univ., Beijing (CN). Key Lab. of Water and Sediment Sciences (Ministry of Education); Sun, Qinghua [Peking Univ., Beijing (CN). Key Lab. of Water and Sediment Sciences (Ministry of Education); Chinese Center for Disease Control and Prevention, Beijing (China). Inst. of Environmental Health and Related Product Safety

2011-05-15

The denitrifier community and associated nitrate and nitrite reduction in the bioaugmented and general sequencing batch reactors (SBRs) during the treatment of coking wastewater containing pyridine and quinoline were investigated. The efficiency and stability of nitrate and nitrite reduction in SBR was considerably improved after inoculation with four pyridine- or quinoline-degrading bacterial strains (including three denitrifying strains). Terminal restriction fragment length polymorphism (T-RFLP) based on the nosZ gene revealed that the structures of the denitrifier communities in bioaugmented and non-bioaugmented reactors were distinct and varied during the course of the experiment. Bioaugmentation protected indigenous denitrifiers from disruptions caused by pyridine and quinoline. Clone library analysis showed that one of the added denitrifiers comprised approximately 6% of the denitrifier population in the bioaugmented sludge. (orig.)

Actions to reduce radioactive emissions: prevention of containment failure by flooding Containment and Reactor Cavity; Acciones para la reduccion de emisiones radiactivas: prevencion del fallo de la Contencion mediante la inundacion de la Contencion y de la Cavidad del Reactor

Energy Technology Data Exchange (ETDEWEB)

Fornos Herrando, J.

2013-07-01

The reactor cavity of Asco and Vandellos II is dry type, thus a severe accident leading to vessel failure might potentially end up resulting in the loss of containment integrity, depending on the viability to cool the molten core. Therefore, significant radioactive emissions could be released to outside. In the framework of Fukushima Stress Tests, ANAV has analyzed the convenience of carrying out different actions to prevent failure of the containment integrity in order to reduce radioactive emissions. The aim of this paper is to present and describe the main phenomenological aspects associated with two of these actions: containment flooding and reactor cavity flooding.

Ageing evaluation model of nuclear reactors structural elements

International Nuclear Information System (INIS)

Ziliukas, A.; Jutas, A.; Leisis, V.

2002-01-01

In this article the estimation of non-failure probability by random faults on the structural elements of nuclear reactors is presented. Ageing is certainly a significant factor in determining the limits of nuclear plant lifetime or life extensions. Usually the non failure probability rates failure intensity, which is characteristic for structural elements ageing in nuclear reactors. In practice the reliability is increased incorrectly because not all failures are fixed and cumulated. Therefore, the methodology with using the fine parameter of the failures flow is described. The comparison of non failure probability and failures flow is carried out. The calculation of these parameters in the practical example is shown too. (author)

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.)

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

Energy Technology Data Exchange (ETDEWEB)

NONE

2013-08-15

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

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

Energy Technology Data Exchange (ETDEWEB)

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

2003-07-01

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

Damage evaluation of 500 MWe Indian pressurized heavy water reactor nuclear containment for air craft impact

International Nuclear Information System (INIS)

Kukreja, Mukesh; Singh, R.K; Vaze, K.K; Kushwaha, H.S.

2003-01-01

Non-linear transient dynamic analysis of 500 MWe Indian Pressurized Heavy Water Reactor (PHWR) nuclear containment has been carried out for the impact of Boeing and Airbus category of aircraft operated in India. The impulsive load time history is generated based on the momentum transfer of the crushable aircraft (soft missiles) of Boeing and Airbus families on the containment structure. The case studies include the analyses of outer containment wall (OCW) single model and the combined model with outer and inner containment wall (ICW) for impulsive loading due to aircraft impact. Initially the load is applied on OCW single model and subsequently the load is transferred to ICW after the local perforation of the OCW is noticed in the transient simulation. In the first stage of the analysis it is demonstrated that the OCW would suffer local perforation with a peak local deformation of 117 mm for impact due to B707-320 and 196 mm due to impact of A300B4 without loss of the overall integrity. However, this first barrier (OCW) cannot absorb the full impulsive load. In the second stage of the analysis of the combined model, the ICW is subjected to lower impulse duration as the load is transferred after 0.19 sec for B707-320 and 0.24 sec for A300B4 due to the local perforation of OCW. This results in the local deformation of approx. 115 mm for B707-320 and 124 mm for A300B4 in ICW and together both the structures (OCW and ICW) are capable of absorbing the full impulsive load. The analysis methodology evolved in the present work would be useful for studying the behaviour of double containment walls and multi barrier structural configurations for aircraft impact with higher energies. The present analysis illustrates that with the provision of double containments for Indian nuclear power plants, adequate reserve strength is available for the case of an extremely low probability event of missile impact generated due commercial aircraft operated in India. (author)

Device for manufacturing methane or synthetic gas from materials containing carbon using a nuclear reactor

International Nuclear Information System (INIS)

Jaeger, W.

1984-01-01

This invention concerns a device for manufacturing methane or synthetic gas from materials containing carbon using a nuclear reactor, where part of the carbon is gasified with hydration and the remaining carbon is converted to synthetic gas by adding steam. This synthetic gas consists mainly of H 2 , CO, CO 2 and CH 4 and can be converted to methane in so-called methanising using a nickel catalyst. The hydrogen gasifier is situated in the first of two helium circuits of a high temperature reactor, and the splitting furnace is situated in the second helium circuit, where part of the methane produced is split into hydrogen at high temperature, which is used for the hydrating splitting of another part of the material containing carbon. (orig./RB) [de

PANDA a multi-purpose thermal-hydraulics facility devoted to nuclear reactor containment safety analysis

International Nuclear Information System (INIS)

Paladino, Domenico

2014-01-01

This paper presents the multi purpose facility PANDA devised for the safety analysis of nuclear reactor containment. The passive safety systems for LWRs have been explained with details about the PAssive Nachzerfallswärmeabfuhr und Druck-Abbau Testanlage (PANDA)

Foreign research reactor spent nuclear fuel inventories containing HEU and LEU of US-origin

International Nuclear Information System (INIS)

Matos, J.E.

1995-01-01

This paper provides estimates of the quantities and types of foreign research reactor spent nuclear fuel containing HEU and LEU of US-origin that are anticipated during the period beginning in January 1996 and extending for 10-15 years

The assessment of structural dynamics problems in nuclear reactor safety

International Nuclear Information System (INIS)

Liebe, R.

1978-10-01

The paper discusses important physical features of structural dynamics problems in reactor safety. First a general characterization is given of the following problems: Containment deformation due to pool-dynamics during BWR-blowdown; behavior of the core internals due to PWR-blowdown loads; dynamic response of a nuclear power plant during an earthquake; fuel element deformation due to local pressure pulses in an LMFBR core. Several criterias are formulated to classify typical problems so that a better choise can be made both of appropriate mathematical/numerical as well as experimental techniques. The degree of physical coupling between structural dynamics and fluid dynamics is discussed in more detail since it requires particular attention when selecting problem-oriented methods of solution. Some examples are given to illustrate the application and to compare advantages and disadvantages of several numerical methods. Then description is given of experimental techniques in structural dynamics and typical problem areas are identified. Finally some results are presented concerning the fuel element deformation problem in LMFBRs and from the general considerations some important conclusions are summarized. (orig.) 891 RW 892 AP [de

Use of molybdenum as a structural material of fuel elements for improving nuclear reactors safety

Energy Technology Data Exchange (ETDEWEB)

Shmelev, Anatoly N.; Kulikov, Gennady G.; Kozhahmet, Bauyrzhan K.; Kulikov, Evgeny G.; Apse, Vladimir A. [National Research Nuclear Univ., Moscow (Russian Federation). Moscow Engineering Physics Institute (MEPhI)

2016-12-15

Main purpose of the study is justifying the use of molybdenum as a structural material of fuel elements for improving the safety of nuclear reactors. Particularity of the used molybdenum is that its isotopic composition corresponds to molybdenum, which is obtained as tailing during operation of the separation cascade for producing a material for medical diagnostics of cancer. The following results were obtained: A method for reducing the thermal constant of fuel elements for light water and fast reactors by using dispersion fuel in cylindrical fuel rods containing, for example, granules of metallic U-Mo-alloy into Mo-matrix was proposed; the necessity of molybdenum enrichment by weakly absorbing isotopes was shown; total use of isotopic molybdenum will be more than 50 %.

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

Energy Technology Data Exchange (ETDEWEB)

Powers, D.A.; Washington, K.E.; Sprung, J.L. [Sandia National Labs., Albuquerque, NM (United States); Burson, S.B. [Nuclear Regulatory Commission, Washington, DC (United States)

1996-07-01

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

Code qualification of structural materials for AFCI advanced recycling reactors.

Energy Technology Data Exchange (ETDEWEB)

Natesan, K.; Li, M.; Majumdar, S.; Nanstad, R.K.; Sham, T.-L. (Nuclear Engineering Division); (ORNL)

2012-05-31

This report summarizes the further findings from the assessments of current status and future needs in code qualification and licensing of reference structural materials and new advanced alloys for advanced recycling reactors (ARRs) in support of Advanced Fuel Cycle Initiative (AFCI). The work is a combined effort between Argonne National Laboratory (ANL) and Oak Ridge National Laboratory (ORNL) with ANL as the technical lead, as part of Advanced Structural Materials Program for AFCI Reactor Campaign. The report is the second deliverable in FY08 (M505011401) under the work package 'Advanced Materials Code Qualification'. The overall objective of the Advanced Materials Code Qualification project is to evaluate key requirements for the ASME Code qualification and the Nuclear Regulatory Commission (NRC) approval of structural materials in support of the design and licensing of the ARR. Advanced materials are a critical element in the development of sodium reactor technologies. Enhanced materials performance not only improves safety margins and provides design flexibility, but also is essential for the economics of future advanced sodium reactors. Code qualification and licensing of advanced materials are prominent needs for developing and implementing advanced sodium reactor technologies. Nuclear structural component design in the U.S. must comply with the ASME Boiler and Pressure Vessel Code Section III (Rules for Construction of Nuclear Facility Components) and the NRC grants the operational license. As the ARR will operate at higher temperatures than the current light water reactors (LWRs), the design of elevated-temperature components must comply with ASME Subsection NH (Class 1 Components in Elevated Temperature Service). However, the NRC has not approved the use of Subsection NH for reactor components, and this puts additional burdens on materials qualification of the ARR. In the past licensing review for the Clinch River Breeder Reactor Project (CRBRP

On elastic structural elements for nuclear reactors

International Nuclear Information System (INIS)

Povolo, F.

1978-03-01

The in-pile stress-relaxation behaviour of materials usually employed for the elastic structural elements, in nuclear reactors, is critically reviewed and the results are compared with those obtained in commercial zirconium alloys irradiated under similar conditions. Finally, it is shown that, under certain conditions, some zirconium alloys may be used as an alternative material for these structural elements. (orig.) [de

Reactor containment depressurization and filtration equipment for use in the case of a serious accident

International Nuclear Information System (INIS)

L'Homme, A.

1987-06-01

A study was carried out under the aegis of the OECD into filtered vented containment systems which permit depressurization of the containment and filtration of the effluents released to the environment, in the event of a major accident with a pressurized water reactor (PWR) (or BWR or CANDU type reactors) involving core meltdown, with a view to minimizing the consequences. This paper describes the various systems examined which could possibly be used for this purpose. These comprised the French robust sand filtration system, the Swedish FILTRA system, the vacuum containment and discharge and emergency filtration system used by the CANDU plants of the Ontario-Hydro electricity company in Canada and the BWR pressure-suppression pounds. The positions of the various national authorities regarding incorporation of such systems into nuclear power plants, the design and technical principles underlying the systems, the procedures and criteria for their use and their advantages and disadvantages are examined [fr

Analysis of aerosol agglomeration and removal mechanisms relevant to a reactor containment

International Nuclear Information System (INIS)

Chiang, H.W.; Mulpuru, S.R.; Lindquist, E.D.

1995-01-01

During some Postulated accidents in a nuclear reactor, radioactive aerosols may be formed and could be released from a rupture of the primary heat transport system into the containment. The released aerosols can agglomerate and form larger aerosol particles. The airborne aerosols can be removed from containment atmosphere by deposition onto the walls and other surfaces in contact with the gas-aerosol mixture. The rate of removal of aerosols depends on the aerosol size, which, in turn, is related to the amount of agglomeration of the aerosol particles. The extent of the removal of the aerosol mass from the containment atmosphere is important in determining the potential radioactive releases to the outside atmosphere. In this paper, selected conditions have been assessed to illustrate the significance of agglomeration for situations potentially of interest in containment safety studies

Method of safely operating nuclear reactor

International Nuclear Information System (INIS)

Ochiai, Kanehiro.

1976-01-01

Purpose: To provide a method of safely operating an nuclear reactor, comprising supporting a load applied to a reactor container partly with secondary container facilities thereby reducing the load borne by the reactor container when water is injected into the core to submerge the core in an emergency. Method: In a reactor emergency, water is injected into the reactor core thereby to submerge the core. Further, water is injected into a gap between the reactor container and the secondary container facilities. By the injection of water into the gap between the reactor container and the secondary container facilities a large apparent mass is applied to the reactor container, as a result of which the reactor container undergoes the same vibration as that of the secondary container facilities. Therefore, the load borne by the reactor container itself is reduced and stress at the bottom part of the reactor container is released. This permits the reactor to be operated more safely. (Moriyama, K.)