The insitu lining of cooling water piping

International Nuclear Information System (INIS)

Vaughan, W.K.; Oxner, K.B.

1994-01-01

The internal corrosion of cooling water piping as well as other industrial piping is becoming an increasing problem to system reliability. There are various alternatives being offered as solutions to the problem including water treatment, coatings, and piping replacement. The in-place lining of these pipes is becoming increasingly popular as a cost-effective method to control corrosion. A cured-in-place plastic composite system can be installed with minimal dismantling or excavation. This paper will examine case histories of the installations of this lining system in power plants at three (3) locations in the United States and one in France. It will also summarize testing that has been performed on the lining system and tests that are currently being performed

Updated pipe break analysis for Advanced Neutron Source Reactor conceptual design

International Nuclear Information System (INIS)

Wendel, M.W.; Chen, N.C.J.; Yoder, G.L.

1994-01-01

The Advanced Neutron Source Reactor (ANSR) is a research reactor to be built at the Oak Ridge National Laboratory that will supply the highest continuous neutron flux levels of any reactor in the world. It uses plate-type fuel with high-mass-flux and highly subcooled heavy water as the primary coolant. The Conceptual Safety Analysis for the ANSR was completed in June 1992. The thermal-hydraulic pipe-break safety analysis (performed with a specialized version of RELAP5/MOD3) focused primarily on double-ended guillotine breaks of the primary piping and some core-damage mitigation options for such an event. Smaller, instantaneous pipe breaks in the cold- and hot-leg piping were also analyzed to a limited extent. Since the initial analysis for the conceptual design was completed, several important changes to the RELAP5 input model have been made reflecting improvements in the fuel grading and changes in the elevation of the primary coolant pumps. Also, a new philosophy for pipe-break safety analysis (similar to that adopted for the New Production Reactor) accentuates instantaneous, limited flow area pipe-break accidents in addition to finite-opening-time, double-ended guillotine breaks of the major coolant piping. This paper discloses the results of the most recent instantaneous pipe-break calculations

Evolution of criteria for repair work on helium lines of Cirus reactor

International Nuclear Information System (INIS)

Mishra, Rajesh; Soni, R.S.; Kushwaha, H.S.

2006-05-01

The research reactor CIRUS uses light water as coolant and heavy water as moderator and is rated for a thermal power of 40 MW. This reactor has been in operation since 1960 and has undergone refurbishment work recently. In the CIRUS reactor, helium gas is utilised as the cover gas. The helium lines are connected with the tube sheet at the top of the calandria. There are eight such helium lines at the top of the calandria, out of which four are connected to one ring header, three to another ring header and the remaining one is single line. These helium gas lines have tongue and groove joints for connecting the stainless steel piping with the aluminium piping. With the prolonged operation of the plant, leakage was observed at these joints. As a part of reactor refurbishing work, these joints were required to be repaired. Since these joints are situated in an inaccessible area, the entire job was to be carried out remotely and therefore, a fail-safe scheme was to be evolved based on computer simulation and analytical work. The entire analysis work had many challenging aspects hence, utmost care was exercised while analytically formulating the scheme for the tightening of these flange joints by postulating the various possible scenarios and by maintaining the stress level within the limits, particularly at the fillet welds between the aluminium pipe and calandria tube sheet. Another challenging aspect of this job was to take care of various uncertainties regarding the prevailing status of the joints. This report highlights the methodology adopted to arrive at the optimum amount of tightening and sequence of tightening. This report also highlights how analytical simulation of actual site scenario was carried out based on site feedbacks at various stages of tightening operations and how strategies were formulated to overcome various challenges and also to take care of various uncertainties in the input information being reported by the site. The tightening work

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

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)

Gap and impact of LMR [Liquid Metal Reactor] piping systems and reactor components

International Nuclear Information System (INIS)

Ma, D.C.; Gvildys, J.; Chang, Y.W.

1987-01-01

Because of high operation temperature, the LMR (Liquid Metal Reactor) plant is characterized by the thin-walled piping and components. Gaps are often present to allow free thermal expansion during normal plant operation. Under dynamic loadings, such as seismic excitation, if the relative displacement between the components exceeds the gap distance, impacts will occur. Since the components and piping become brittle over their design lifetime, impact is of important concern for it may lead to fractures of components and other serious effects. This paper deals with gap and impact problems in the LMR reactor components and piping systems. Emphasis is on the impacts due to seismic motion. Eight sections are contained in this paper. The gap and impact problems in LMR piping systems are described and a parametric study is performed on the effects of gap-induced support nonlinearity on the dynamics characteristics of the LMR piping systems. Gap and impact problems in the LMR reactor components are identified and their mathematical models are illustrated, and the gap and impact problems in the seismic reactor scram are discussed. The mathematical treatments of various impact models are also described. The uncertainties in the current seismic impact analyses of LMR components and structures are presented. An impact test on a 1/10-scale LMR thermal liner is described. The test results indicated that several clusters of natural modes can be excited by the impact force. The frequency content of the excited modes depends on the duration of the impact force; the shorter the duration, the higher the frequency content

SCC-induced failure of a 304 stainless steel pipe

International Nuclear Information System (INIS)

Tapping, R.L.; Disney, D.J.; Szostak, F.J.

1993-01-01

On 1991 January 12, a 304 Stainless Steel (SS) suction line in the AECL-Research NRU reactor failed, shutting down the reactor for approximately 12 months. The pipe, a 32 mm schedule 40 304 stainless steel line exposed to D 2 O at temperatures ≤35 degrees C had been in service for approximately 20 years, although no manufacturing data or composition specifications were available. The failure and resultant leak resulted in a small loss of D 2 O moderator from the reactor vessel. The pipe cracked approximately 180 degrees C around the circumference of a weld. This failure was unexpected and hense a thorough metallographic examination was carried out on the failed section, on the rest of the line (Line 1212), and on representative samples from the rest of the reactor in order to assess the integrity of the remaining piping

Tentative design-philosophy for bellows in sodium cooled fast breeder reactors pipings

Energy Technology Data Exchange (ETDEWEB)

Scaller, K; Vrillon, B

1980-02-01

Expansion joints have proved to be reliable components, when properly designed and realized, in normal industrial equipment. But nevertheless bellows have not been employed widely in nuclear reactors and almost not in sodium cooled fast breeder reactors, where use of expansion-joints could considerably shorten the length of pipelines and, in consequence, lower the cost of the power plant. In the framework of its research and development program on fast reactors the French Atomic Energy.Commission, in cooperation with the industry, develops guidelines, backed up by experiments, to allow a safe design of pipe-lines and compensating-devices. The main points of these guidelines are discussed in this paper with the understanding, that they are tentative rules subject to changes. The guidelines are a complement to existing rules, like ASME - Code III, Code Case 1481, standards of the EJMA Preliminary Draft for Code Case Class I, Expansion Joints in Piping systems and suppliers' rules for the special case of application to sodium cooled fast breeder reactors. Relatively small diameters and easily accessible expansion joints, on control rods and valves for example, are not concerned. These guidelines do not apply to the bellows which are used as an integral part of a component.

Tentative design-philosophy for bellows in sodium cooled fast breeder reactors pipings

International Nuclear Information System (INIS)

Scaller, K.; Vrillon, B.

1980-01-01

Expansion joints have proved to be reliable components, when properly designed and realized, in normal industrial equipment. But nevertheless bellows have not been employed widely in nuclear reactors and almost not in sodium cooled fast breeder reactors, where use of expansion-joints could considerably shorten the length of pipelines and, in consequence, lower the cost of the power plant. In the framework of its research and development program on fast reactors the French Atomic Energy.Commission, in cooperation with the industry, develops guidelines, backed up by experiments, to allow a safe design of pipe-lines and compensating-devices. The main points of these guidelines are discussed in this paper with the understanding, that they are tentative rules subject to changes. The guidelines are a complement to existing rules, like ASME - Code III, Code Case 1481, standards of the EJMA Preliminary Draft for Code Case Class I, Expansion Joints in Piping systems and suppliers' rules for the special case of application to sodium cooled fast breeder reactors. Relatively small diameters and easily accessible expansion joints, on control rods and valves for example, are not concerned. These guidelines do not apply to the bellows which are used as an integral part of a component

Study on concept of web-based reactor piping design data platform

International Nuclear Information System (INIS)

Wang Yu; Zhou Yu; Dong Jianling; Meng Yang

2005-01-01

For solving the piping design problems such as design data deficiency, designer communication inconvenience and design project inconsistence, Reactor Piping Design Database Platform, which is the main part of the Integrated Nuclear Project Research Platform, is proposed by analyzing the nuclear piping designs in detail. The functions and system structures of the platform are described in the paper for the sake of the realization of the Reactor Piping Design Database Platform. The platform is constituted by web-based management interface, AutoPlant selected as CAD software, and relation database management system (DBMS). (authors)

78 FR 62614 - Guttman Energy, Inc., PBF Holding Company LLC v. Buckeye Pipe Line Company, L.P., Laurel Pipe...

Science.gov (United States)

2013-10-22

... DEPARTMENT OF ENERGY Federal Energy Regulatory Commission [Docket No. OR14-4-000] Guttman Energy, Inc., PBF Holding Company LLC v. Buckeye Pipe Line Company, L.P., Laurel Pipe Line Company, L.P... complaint against Buckeye Pipe Line Company L.P. and Laurel Pipe Line Company L.P. (Respondents) challenging...

Reactor process water (PW) piping inspections, 1984--1990

International Nuclear Information System (INIS)

Ehrhart, W.S.; Elder, J.B.; Sprayberry, R.E.; Vande Kamp, R.W.

1990-01-01

In July 1983, the NRC ordered the shutdown of five boiling water reactors (BWR's) because of concerns about reliability of ultrasonic examination for detecting intergranular stress corrosion cracking (IGSCC). These concerns arose because of leaking piping at Niagara Mohawk's Nine Mile Point which was attributed to IGSCC. The leaks were detected shortly after completion of ultrasonic examinations of the piping. At that time, the Dupont plant manager at Savannah River (SR) directed that investigations be performed to determine if similar problems could exist in SR reactors. Investigation determined that all conditions believed necessary for the initiation and propagation of IGSCC in austenitic stainless steel exist in SR reactor process water (PW) systems. Sensitized, high carbon, austenitic stainless steel, a high purity water system with high levels of dissolved oxygen, and the residual stresses associated with welding during construction combine to provide the necessary conditions. A periodic UT inspection program is now in place to monitor the condition of the reactor PW piping systems. The program is patterned after NRC NUREG 0313, i.e., welds are placed in categories based on their history. Welds in upgraded or replacement piping are examined on a standard schedule (at least every five years) while welds with evidence of IGSCC, evaluated as acceptable for service, are inspected at every extended outage (15 to 18 months). This includes all welds in PW systems three inches in diameter and above. Welds are replaced when MSCC exceeds the replacement criteria of more than twenty percent of pipe circumference of fifty percent of through-wall depth. In the future, we intend to perform flow sizing with automated UT techniques in addition to manual sizing to provide more information for comparison with future examinations

The installation welding of pressure water reactor coolant piping

International Nuclear Information System (INIS)

Deng Feng

2010-01-01

Large pressure water reactor nuclear power plants are constructing in our country. There are three symmetry standard loops in reactor coolant system. Each loop possesses a steam generator and a primary poop, in which one of the loops is equipped with a pressurizer. These components are connected with reactor pressure vessel by installation welding of the coolant piping. The integrity of reactor coolant pressure boundary is the second barrier to protect the radioactive substance from release to outside, so the safe operation of nuclear power plant is closely related to the quality of coolant piping installation welding. The heavy tube with super low carbon content austenitic stainless steel is selected for coolant piping. This kind of material has good welding behavior, but the poor thermal conductivity, the big liner expansion coefficient and the big welding deformation will cause bigger welding stress. To reduce the welding deformation, to control the dimension precision, to reduce the residual stress and to ensure the welding quality the installation sequence should be properly designed and the welding technology should be properly controlled. (authors)

Shielding computations for solution transfer lines from Analytical Lab to process cells of Demonstration Fast Reactor Plant (DFRP)

International Nuclear Information System (INIS)

Baskar, S.; Jose, M.T.; Baskaran, R.; Venkatraman, B.

2018-01-01

The diluted virgin solutions (both aqueous and organic) and aqueous analytical waste generated from experimental analysis of process solutions, pertaining to Fast Breeder Test Reactor (FBTR) and Prototype Fast Breeder Reactor (PFBR), in glove boxes of active analytical Laboratory (AAL) are pumped back to the process cells through a pipe in pipe arrangement. There are 6 transfer lines (Length 15-32 m), 2 for each type of transfer. The transfer lines passes through the area inside the AAL and also the operating area. Hence it is required to compute the necessary radial shielding requirement around the lines to limit the dose rates in both the areas to the permissible values as per the regulatory requirement

Fracture assessment of Savannah River Reactor carbon steel piping

International Nuclear Information System (INIS)

Mertz, G.E.; Stoner, K.J.; Caskey, G.R.; Begley, J.A.

1991-01-01

The Savannah River Site (SRS) production reactors have been in operation since the mid-1950's. One postulated failure mechanism for the reactor piping is brittle fracture of the original A285 and A53 carbon steel piping. Material testing of archival piping determined (1) the static and dynamic tensile properties; (2) Charpy impact toughness; and (3) the static and dynamic compact tension fracture toughness properties. The nil-ductility transition temperature (NDTT), determined by Charpy impact test, is above the minimum operating temperature for some of the piping materials. A fracture assessment was performed to demonstrate that potential flaws are stable under upset loading conditions and minimum operating temperatures. A review of potential degradation mechanisms and plant operating history identified weld defects as the most likely crack initiation site for brittle fracture. Piping weld defects, as characterized by radiographic and metallographic examination, and low fracture toughness material properties were postulated at high stress locations in the piping. Normal operating loads, upset loads, and residual stresses were assumed to act on the postulated flaws. Calculated allowable flaw lengths exceed the size of observed weld defects, indicating adequate margins of safety against brittle fracture. Thus, a detailed fracture assessment was able to demonstrate that the piping systems will not fail by brittle fracture, even though the NDTT for some of the piping is above the minimum system operating temperature

Seismic fragility analysis of buried steel piping at P, L, and K reactors

International Nuclear Information System (INIS)

Wingo, H.E.

1989-10-01

Analysis of seismic strength of buried cooling water piping in reactor areas is necessary to evaluate the risk of reactor operation because seismic events could damage these buried pipes and cause loss of coolant accidents. This report documents analysis of the ability of this piping to withstand the combined effects of the propagation of seismic waves, the possibility that the piping may not behave in a completely ductile fashion, and the distortions caused by relative displacements of structures connected to the piping

Investigation and evaluation of cracking incidents in piping in pressurized water reactors. Technical report

International Nuclear Information System (INIS)

1980-09-01

This report summarizes an investigation of known cracking incidents in pressurized water reactor plants. Several instances of cracking in feedwater piping in 1979, together with reported cases of stress corrosion cracking at Three Mile Island Unit 1, led to the establishment of the third Pipe Crack Study Group. Major differences between the scope of the third PCSG and the previous two are: (1) the emphasis given to systems safety implications of cracking, and (2) the consideration given all cracking mechanisms known to affect PWR piping, including the failure of small lines in secondary safety systems. The present PCSG reviewed existing information on cracking of PWR pipe systems, either contained in written records of collected from meetings in the United States, and made recommendations in response to the PCSG charter questions and to othe major items that may be considered to either reduce the potential for cracking or to improve licensing bases

Evaluation of the influence of seismic restraint characteristics on breeder reactor piping systems

International Nuclear Information System (INIS)

Mello, R.M.; Pollono, L.P.

1979-01-01

For the Clinch River Breeder Reactor Plant (CRBRP) heat transport system piping within the reactor containment building, dynamic analyses of the piping loops have been performed to study the effect of restraint stiffness on the dynamic behavior of the piping. In addition, analysis and testing of typical CRBRP restraint system components have been performed for the purpose of quantifying and verifying the basic characteristics of the restraints used in the piping system dynamic analysis

Development of integrated insulation joint for cooling pipe in tokamak reactor

International Nuclear Information System (INIS)

Nishio, Satoshi; Abe, Tetsuya; Kawamura, Masashi; Yamazaki, Seiichiro.

1994-08-01

In a tokamak fusion reactor, an electrically insulated part is needed for an in-vessel piping system in order to break an electric circuit loop. When a closed loop is formed in the piping system, large induced electromagnetic forces during a plasma disruption (rapid plasma current quench) could give damages on the piping system. Ceramic brazing joint is a conventional method for the electric circuit break, but an application to the fusion reactor is not feasible due to its brittleness. Here, a stainless steel/ceramics/stainless steel functionally gradient material (FGM) has been proposed and developed as an integrated insulation joint of the piping system. Both sides of the joint can be welded to the main pipes, and expected to be reliable even in the fusion reactor environment. When the FGM joint is manufactured by way of a sintering process, a residual thermal stress is the key issue. Through detailed computations of the residual thermal stress and several trial productions, tubular elements of FGM joints have been successfully manufactured. (author)

Leak test of the pipe line for radioactive liquid waste

International Nuclear Information System (INIS)

Machida, Chuji; Mori, Shoji.

1976-01-01

In the Tokai Research Establishment, most of the radioactive liquid waste is transferred to a wastes treatment facility through pipe lines. As part of the pipe lines a cast iron pipe for town gas is used. Leak test has been performed on all joints of the lines. For the joints buried underground, the test was made by radioactivity measurement of the soil; and for the joints in drainage ditch by the pressure and bubble methods. There were no leakage at all, indicating integrity of all the joints. On the other hand, it is also known by the other test that the corrosion of inner surface of the piping due to liquid waste is only slight. The pipe lines for transferring radioactive liquid waste are thus still usable. (auth.)

Pipe Lines – External Corrosion

Directory of Open Access Journals (Sweden)

Dan Babor

2008-01-01

Full Text Available Two areas of corrosion occur in pipe lines: corrosion from the medium carried inside the pipes; corrosion attack upon the outside of the pipes (underground corrosion. Electrolytic processes are also involved in underground corrosion. Here the moisture content of the soil acts as an electrolyte, and the ions required to conduct the current are supplied by water-soluble salts (chlorides, sulfates, etc. present in the soil. The nature and amount of these soluble materials can vary within a wide range, which is seen from the varying electrical conductivity and pH (varies between 3 and 10. Therefore the characteristics of a soil will be an important factor in under-ground corrosion.

Steam line break analysis in CAREM-25 reactor

International Nuclear Information System (INIS)

Zanocco, Pablo; Gimenez, Marcelo O.; Vertullo, Alicia; Schlamp, Miguel A.; Garcia, Alicia E.

2000-01-01

The main objective of this report is to analyze the reactor response during a steam line break postulated accident with RELAP5, a plant code using a separated flow model. The steam line break caused a rapid blowdown of the secondary coolant increasing the heat removal in the steam generator. As a consequence and due to reactor features the core power is also increased. As maximum removed power in the secondary side is highly dependant on the total water volume evaporated during the accident a detailed model of feed water and outlet steam pipes is provided. Different cases are analyzed with and without feedwater system and considering the fail or success of the First Shutdown System. In all the sequences the DNBR and CPR remain above the minimum safety values established by design. Further calculations concerning depressurization ramps and steam generator feed water pumps response during depressurization are advised. (author)

Steam Line Break Analysis in CAREM-25 Reactor

International Nuclear Information System (INIS)

Zanocco, Pablo; Gimenez, Marcelo; Vertullo, Alicia; Garcia, A; Schlamp, Miguel

2000-01-01

The main objective of this report is to analyze the reactor response during a steam line break postulated accident with RELAP5, a plant code using a separated flow model.The steam line break caused a rapid blowdown of the secondary coolant increasing the heat removal in the steam generator.As a consequence and due to reactor features the core power is also increased.As maximum removed power in the secondary side is highly dependant on the total water volume evaporated during the accident a detailed model of feed water and outlet steam pipes is provided.Different cases are analyzed with and without feedwater system and considering the fail or success of the First Shutdown System.In all the sequences the DNBR and CPR remain above the minimum safety values established by design.Further calculations concerning depressurization ramps and steam generator feed water pumps response during depressurization are advised

78 FR 60897 - Certain Welded Large Diameter Line Pipe From Japan

Science.gov (United States)

2013-10-02

... Diameter Line Pipe From Japan Determination On the basis of the record \\1\\ developed in the subject five... order on certain welded large diameter line pipe from Japan would likely to lead to continuation or... Line Pipe from Japan: Investigation No. 731-TA-919 (Second Review). By order of the Commission. Issued...

Investigation and evaluation of stress-corrosion cracking in piping of light water reactor plants

International Nuclear Information System (INIS)

1979-01-01

In 1975, a Pipe Cracking Study Group, established by the United States Nuclear Regulatory Commission (USNRC), reviewed intergranular stress-corrosion cracking (IGSCC) in Bioling Water Reactors (BWRs) and issued a report. During 1978, IGSCC was reported for the first time in large-diameter piping (> 20 in.) in a BWR in Germany. This discovery, together with the reported questions concerning the interpretation of ultrasonic inspections, led to the activation of a new Pipe Crack Study Group (PCSG) by USNRC. The charter of the new PCSG was expanded: (1) to include review of potential for stress-corrosion cracking in Pressurized Water Reactors (PWRs) as well as BWRs, (2) to examine operating experience in foreign reactors relevant to IGSCC, and (3) to study five specific questions. The PCSG limited the scope of the study to BWR and PWR piping runs and safe ends attached to the reactor pressure vessel. Not considered were components such as the reactor pressure vessel, pumps, valves, steam generators, large steam turbines, etc. Throughout this report, as well as in the title, the safe ends are arbitrarily defined as piping

Pressurized water-reactor feedwater piping response to water hammer

International Nuclear Information System (INIS)

Arthur, D.

1978-03-01

The nuclear power industry is interested in steam-generator water hammer because it has damaged the piping and components at pressurized water reactors (PWRs). Water hammer arises when rapid steam condensation in the steam-generator feedwater inlet of a PWR causes depressurization, water-slug acceleration, and slug impact at the nearest pipe elbow. The resulting pressure pulse causes the pipe system to shake, sometimes violently. The objective of this study is to evaluate the potential structural effects of steam-generator water hammer on feedwater piping. This was accomplished by finite-element computation of the response of two sections of a typical feedwater pipe system to four representative water-hammer pulses. All four pulses produced high shear and bending stresses in both sections of pipe. Maximum calculated pipe stresses varied because the sections had different characteristics and were sensitive to boundary-condition modeling

Nonlinear dynamic analysis of high energy line pipe whip

International Nuclear Information System (INIS)

Hsu, L.C.; Kuo, A.Y.; Tang, H.T.

1983-01-01

To facilitate potential cost savings in pipe whip protection design, TVA conducted a 1'' high pressure line break test to investigate the pipe whip behavior. The test results are available to EPRI as a data base for a generic study on nonlinear dynamic behavior of piping systems and pipe whip phenomena. This paper describes a nonlinear dynamic analysis of the TVA high energy line tests using ABAQUS-EPGEN code. The analysis considers the effects of large deformation and high strain rate on resisting moment and energy absorption capability of the analyzed piping system. The numerical results of impact forces, impact velocities, and reaction forces at pipe supports are compared to the TVA test data. The pipe whip impact time and forces have also been calculated per the current NRC guidelines and compared. The calculated pipe support reaction forces prior to impact have been found to be in good agreement with the TVA test data except for some peak values at the very beginning of the pipe break. These peaks are believed to be due to stress wave propagation which cannot be addressed by the ABAQUS code. Both the effects of elbow crushing and strain rate have been approximately simulated. The results are found to be important on pipe whip impact evaluation. (orig.)

Failure rates in Barsebaeck-1 reactor coolant pressure boundary piping. An application of a piping failure database

International Nuclear Information System (INIS)

Lydell, B.

1999-05-01

This report documents an application of a piping failure database to estimate the frequency of leak and rupture in reactor coolant pressure boundary piping. The study used Barsebaeck-1 as reference plant. The study tried two different approaches to piping failure rate estimation: 1) PSA-style, simple estimation using Bayesian statistics, and 2) fitting of statistical distribution to failure data. A large, validated database on piping failures (like the SKI-PIPE database) supports both approaches. In addition to documenting leak and rupture frequencies, the SKI report describes the use of piping failure data to estimate frequency of medium and large loss of coolant accidents (LOCAs). This application study was co sponsored by Barsebaeck Kraft AB and SKI Research

Analysis of a postulated pipe rupture and subsequent check valve slam of a PWR feedwater line

International Nuclear Information System (INIS)

Chang, K.C.; Adams, T.M.

1983-01-01

System designs criteria employed in the design of pressurized water reactors (PWR) requires that, for a postulated instantaneous guillotine rupture anywhere in the steam generator feedwater system, no more than one steam generator can be allowed to blowdown. Feedwater systems in many PWR's consist of pipe lines running from the feedwater pumps into a common feedwater header then branching into each steam generator from the header. The feedwater piping to each steam generator contains swing check valves to prevent reverse flow from the steam generator. This activation of some or all of these check valves significantly complicates the system structural analysis in that not only the blowdown forces resulting from the postulated pipe rupture, but also the water hammer loads resulting from closure of the check valve at high reverse flow velocities must be considered. The loads resulting from system blowdown and check valve closure are axial in nature. Peak loads ranging from 130000 lbs. to 180000 lbs. are not uncommon and are layout dependent. The analysis and design to withstand this transient loading deviates from the usual feedwater line design in that supports are required along the piping axis in the direction normal to the usual seismic supports. A brief and general discussion of the methods employed in the generation of the thermal-hydraulic loadings is presented. However, the discussion emphasizes the piping and piping support structural design and analysis method and approaches used in evaluating a selected portion of such a feedwater system. (orig./RW)

Failure rates in Barsebaeck-1 reactor coolant pressure boundary piping. An application of a piping failure database

Energy Technology Data Exchange (ETDEWEB)

Lydell, B. [RSA Technologies, Vista, CA (United States)

1999-05-01

This report documents an application of a piping failure database to estimate the frequency of leak and rupture in reactor coolant pressure boundary piping. The study used Barsebaeck-1 as reference plant. The study tried two different approaches to piping failure rate estimation: 1) PSA-style, simple estimation using Bayesian statistics, and 2) fitting of statistical distribution to failure data. A large, validated database on piping failures (like the SKI-PIPE database) supports both approaches. In addition to documenting leak and rupture frequencies, the SKI report describes the use of piping failure data to estimate frequency of medium and large loss of coolant accidents (LOCAs). This application study was co sponsored by Barsebaeck Kraft AB and SKI Research 41 refs, figs, tabs

Boiling water reactor

International Nuclear Information System (INIS)

Matsumoto, Tomoyuki; Inoue, Kotaro; Ishida, Masayoshi.

1975-01-01

Object: To connect a feedwater pipe to a recycling pipe line, the recycling pipe line being made smaller in diameter, thereby minimizing loss of coolant resulting from rupture of the pipe and improving safety against trouble of coolant loss. Structure: A feedwater pipe is directly connected to a recycling pipe line before a booster pump, and a mixture of recycling water and feedwater is increased in pressure by the booster pump, after which it is introduced into a jet pump in the form of water for driving the jet pump to suck surrounding water causing it to be flown into the core. In accordance with the abovementioned structure, since the flow of feedwater can be used as a part of water for driving the jet pump, the flow within the recycling pipe line may be decreased so that the recycling pipe line can be made smaller in diameter to reduce the flow of coolant in the reactor, which flows out when the pipe is ruptured. (Furukawa, Y.)

Piping benchmark problems for the General Electric Advanced Boiling Water Reactor

International Nuclear Information System (INIS)

Bezler, P.; DeGrassi, G.; Braverman, J.; Wang, Y.K.

1993-08-01

To satisfy the need for verification of the computer programs and modeling techniques that will be used to perform the final piping analyses for an advanced boiling water reactor standard design, three benchmark problems were developed. The problems are representative piping systems subjected to representative dynamic loads with solutions developed using the methods being proposed for analysis for the advanced reactor standard design. It will be required that the combined license holders demonstrate that their solutions to these problems are in agreement with the benchmark problem set

Analytical prediction on the pump-induced pulsating pressure in a reactor coolant pipe

International Nuclear Information System (INIS)

Lee, K.B.; Im, I.Y.; Lee, S.K.

1992-01-01

An analytical method is presented for predicting the amplitudes of pump-induced fluctuating pressures in a reactor coolant pipe using a linear transformation technique which reduces a homogeneous differential equation with non-homogeneous boundary conditions into a nonhomogeneous differential equation with homogeneous boundary conditions. At the end of the pipe, three types of boundary conditions are considered-open, closed and piston-spring supported. Numerical examples are given for a typical reactor. Comparisons of measured pressure amplitudes in the pipe with model prediction are shown to be in good agreement for the forcing frequencies. (author)

Study on unstable fracture characteristics of light water reactor piping

International Nuclear Information System (INIS)

Kurihara, Ryoichi

1998-08-01

Many testing studies have been conducted to validate the applicability of the leak before break (LBB) concept for the light water reactor piping in the world. It is especially important among them to clarify the condition that an inside surface crack of the piping wall does not cause an unstable fracture but ends in a stable fracture propagating only in the pipe thickness direction, even if the excessive loading works to the pipe. Pipe unstable fracture tests performed in Japan Atomic Energy Research Institute had been planned under such background, and clarified the condition for the cracked pipe to cause the unstable fracture under monotonous increase loading or cyclic loading by using test pipes with the inside circumferential surface crack. This paper examines the pipe unstable fracture by dividing it into two parts. One is the static unstable fracture that breaks the pipe with the inside circumferential surface crack by increasing load monotonously. Another is the dynamic unstable fracture that breaks the pipe by the cyclic loading. (author). 79 refs

Development of prototype reactor maintenance. (2) Application to piping support of sodium-cooled reactor prototype

International Nuclear Information System (INIS)

Arai, Masanobu; Kunogi, Kosuke; Aizawa, Kosuke; Chikazawa, Yoshitaka; Takaya, Shigeru; Kubo, Shigenobu; Kotake, Shoji; Ito, Takaya; Yamaguchi, Akira

2017-01-01

A maintenance program on piping support of prototype fast breeder reactor Monju are studied. Based on degradation mechanism, snubbers in Monju primary cooling system showed lifetime more than the plant lifetime of 30 years by experiments conservatively. For the first step during construction, visual inspection on accessible all supports could be available. In that visual inspection, mounting conditions and damages of all accessible supports could be monitored. One of major features of the Monju primary piping system is large thermal expansion due to large temperature difference between maintenance and operation conditions. Thanks to that large thermal expansion, integrity of piping supports could be monitored by measuring piping displacement. When technologies of piping displacement monitoring are matured in Monju, visual inspection on piping support could be shifted to piping displacement monitoring. At that stage, the visual inspection could be limited only on representative supports. (author)

Repair and preventive maintenance technology for BWR reactor internals and piping

International Nuclear Information System (INIS)

Ootsubo, Tooru; Itou, Takashi; Sakashita, Akihiro

2009-01-01

Stress corrosion cracking of welding portion has found in many domestic and foreign BWR reactor internals and Primary Loop Recirculation piping. Also, repair and preventive maintenance technologies for SCC has been developed and/or adopted to BWRs in recent years. This paper introduces the sample of these technologies, such as seal-welding for SCC on BWR reactor internals, preventive maintenance technology for PLR piping such as Corrosion Resistant Cladding, Internal Polishing and Induction Heating Stress Improvement. These technologies are introduced on 'E-Journal of Advanced Maintenance', which is an international journal on a exclusive website of Japan Society of Maintenology. (author)

Reactor shutdown device

International Nuclear Information System (INIS)

Inoue, Toyokazu.

1982-01-01

Purpose: To obtain a highly reliable reactor shutdown device capable of checking its function irrespective of the state whether shutdown or operation in a gas-cooled type reactor. Constitution: A hopper is disposed above a guide tube inserted into the reactor core and particulate neutron absorbers are contained in the hopper. An opening for falling particles is disposed to the bottom of the hopper in opposition to the upper end of the guide pipe and the opening is closed by a plug suspended by way of a weld line so as to be capable of dropping. A power source for supplying electrical current to the weld line is disposed. Accordingly, if the current is supplied to the weld line, the line is cut by welding to fall the plug so that the neutron-absorbing particles fall from the opening into the guide pipe to shutdown the reactor, whereby high reliability is obtained for the operation. (Seki, T.)

Seismic analysis of liquid metal reactor piping systems

International Nuclear Information System (INIS)

Wang, C.Y.

1987-01-01

To safely assess the adequacy of the LMR piping, a three-dimensional piping code, SHAPS, has been developed at Argonne National Laboratory. This code was initially intended for calculating hydrodynamic-wave propagation in a complex piping network. It has salient features for treating fluid transients of fluid-structure interactions for piping with in-line components. The code also provides excellent structural capabilities of computing stresses arising from internal pressurization and 3-D flexural motion of the piping system. As part of the development effort, the SHAPS code has been further augmented recently by introducing the capabilities of calculating piping response subjected to seismic excitations. This paper describes the finite-element numerical algorithm and its applications to LMR piping under seismic excitations. A time-history analysis technique using the implicit temporal integration scheme is addressed. A 3-D pipe element is formulated which has eight degrees of freedom per node (three displacements, three rotations, one membrane displacement, and one bending rotation) to account for the hoop, flexural, rotational, and torsional modes of the piping system. Both geometric and material nonlinearities are considered. This algorithm is unconditionally stable and is particularly suited for the seismic analysis

Accident analysis of heat pipe cooled and AMTEC conversion space reactor system

International Nuclear Information System (INIS)

Yuan, Yuan; Shan, Jianqiang; Zhang, Bin; Gou, Junli; Bo, Zhang; Lu, Tianyu; Ge, Li; Yang, Zijiang

2016-01-01

Highlights: • A transient analysis code TAPIRS for HPS has been developed. • Three typical accidents are analyzed using TAPIRS. • The reactor system has the self-stabilization ability under accident conditions. - Abstract: A space power with high power density, light weight, low cost and high reliability is of crucial importance to future exploration of deep space. Space reactor is an excellent candidate because of its unique characteristics of high specific power, low cost, strong environment adaptability and so on. Among all types of space reactors, heat pipe cooled space reactor, which adopts the passive heat pipe (HP) as core cooling component, is considered as one of the most promising choices and is widely studied all over the world. This paper develops a transient analysis code (TAPIRS) for heat pipe cooled space reactor power system (HPS) based on point reactor kinetics model, lumped parameter core heat transfer model, combined HP model (self-diffusion model, flat-front startup model and network model), energy conversion model of Alkali Metal Thermal-to-Electric Conversion units (AMTEC), and HP radiator model. Three typical accidents, i.e., control drum failure, AMTEC failure and partial loss of the heat transfer area of radiator are then analyzed using TAPIRS. By comparing the simulation results of the models and steady state with those in the references, the rationality of the models and the solution method is validated. The results show the following. (1) After the failure of one set of control drums, the reactor power finally reaches a stable value after two local peaks under the temperature feedback. The fuel temperature rises rapidly, however it is still under safe limit. (2) The fuel temperature is below a safe limit under the AMTEC failure and partial loss of the heat transfer area of radiator. This demonstrates the rationality of the system design and the potential applicability of the TAPIRS code for the future engineering application of

Solid0Core Heat-Pipe Nuclear Batterly Type Reactor

Energy Technology Data Exchange (ETDEWEB)

Ehud Greenspan

2008-09-30

This project was devoted to a preliminary assessment of the feasibility of designing an Encapsulated Nuclear Heat Source (ENHS) reactor to have a solid core from which heat is removed by liquid-metal heat pipes (HP).

Reactor Primary Coolant System Pipe Rupture Study. Progress report No. 32, July--December 1974

International Nuclear Information System (INIS)

1975-03-01

The pipe rupture study is designed to extend the understanding of failure-causing mechanisms and to provide improved capability for evaluating reactor piping systems to minimize the probability of failures. Following a detailed review to determine the effort most needed to improve nuclear system piping (Phase I), analytical and experimental efforts (Phase II) were started in 1965. This progress report summarizes the recent accomplishments of a broad program in (a) basic fatigue studies focused on Elastic/Plastic ASME Code Design Rules, (b) at-reactor tests of the effect of primary coolant environment on the fatigue behavior of piping steels, and (c) studies directed at quantifying weld sensitization in T-304 stainless steel. (auth)

78 FR 69078 - Houston Pipe Line Company LP; Notice of Application

Science.gov (United States)

2013-11-18

... DEPARTMENT OF ENERGY Federal Energy Regulatory Commission [Docket No. CP14-13-000] Houston Pipe Line Company LP; Notice of Application Take notice that on October 28, 2013, Houston Pipe Line Company LP (HPL), 1300 Main Street, Houston, Texas 77002, filed an application in Docket No. CP14-13-000...

Underground nuclear power station using self-regulating heat-pipe controlled reactors

International Nuclear Information System (INIS)

Hampel, V.E.

1989-01-01

The author presents a nuclear reactor for generating electricity disposed underground at the bottom of a vertical hole that can be drilled using conventional drilling technology. The primary coolant of the reactor core is the working fluid in a plurality of thermodynamically coupled heat pipes emplaced in the hole between the heat source at the bottom of the hole and heat exchange means near the surface of the earth. Additionally, the primary coolant (consisting of the working fluid in the heat pipes in the reactor core) moderates neutrons and regulates their reactivity, thus keeping the power of the reactor substantially constant. At the end of its useful life, the reactor core may be abandoned in place. Isolation from the atmosphere in case of accident or for abandonment is provided by the operation of explosive closures and mechanical valves emplaced along the hole. This invention combines technology developed and tested for small, highly efficient, space-based nuclear electric power plants with the technology of fast-acting closure mechanisms developed and used for underground testing of nuclear weapons. This invention provides a nuclear power installation which is safe from the worst conceivable reactor accident, namely, the explosion of a nuclear weapon near the ground surface of a nuclear power reactor

Underground nuclear power station using self-regulating heat-pipe controlled reactors

Science.gov (United States)

Hampel, Viktor E.

1989-01-01

A nuclear reactor for generating electricity is disposed underground at the bottom of a vertical hole that can be drilled using conventional drilling technology. The primary coolant of the reactor core is the working fluid in a plurality of thermodynamically coupled heat pipes emplaced in the hole between the heat source at the bottom of the hole and heat exchange means near the surface of the earth. Additionally, the primary coolant (consisting of the working flud in the heat pipes in the reactor core) moderates neutrons and regulates their reactivity, thus keeping the power of the reactor substantially constant. At the end of its useful life, the reactor core may be abandoned in place. Isolation from the atmosphere in case of accident or for abandonment is provided by the operation of explosive closures and mechanical valves emplaced along the hole. This invention combines technology developed and tested for small, highly efficient, space-based nuclear electric power plants with the technology of fast-acting closure mechanisms developed and used for underground testing of nuclear weapons. This invention provides a nuclear power installation which is safe from the worst conceivable reactor accident, namely, the explosion of a nuclear weapon near the ground surface of a nuclear power reactor.

Acoustical holographic Siamese image technique for imaging radial cracks in reactor piping

International Nuclear Information System (INIS)

Collins, H.D.; Gribble, R.P.

1985-04-01

This paper describes a unique technique (i.e., ''Siamese imaging'') for imaging quasi-vertical defects in reactor pipe weldments. The Siamese image is a bi-symmetrical view of the inner surface defect. Image construction geometry consists of two probes (i.e., source/receiver) operating either from opposite sides or the same side of the defect to be imaged. As the probes are scanned across a lower surface connected defect, they encounter two images - first the normal upright image and then the inverted image. The final integrated image consists of two images connected along their baselines, thus we call it a ''Siamese image.'' The experimental imaging results on simulated and natural cracks in reactor piping weldments graphically illustrate this unique technique. Excellent images of mechanical fatique and thermal cracks were obtained on ferritic and austenitic piping

Ultrasonic testing with the phased array method at the pipe connection inner edges in pipings

International Nuclear Information System (INIS)

Brekow, G.; Wuestenberg, H.; Hesselmann, H.; Rathgeb, W.

1991-01-01

Ultrasonic testing with the phased array method at the pipe connection inner edges in pipings. The pipe connection inner corner tests in feedwater lines to the main coolant pipe were carried out by Preussen-Elektra in cooperation with Siemens KWU and the BAM with the ultrasonic phased array method. The testing plan was developed by means of a computed model. For a trial of the testing plan, numerous ultrasonic measurements with the phased array method were carried out using a pipe test piece with TH-type inner edges, which was a 1:1 model of the reactor component to be tested. The data measured at several test notches in the pipe connection inner edge area covered by a plating of 6 mm were analyzed. (orig./MM) [de

WATER QUALITY AND TREATMENT CONSIDERATIONS FOR CEMENT-LINED AND A-C PIPE

Science.gov (United States)

Both cement mortar lined (CML) and asbestos-cement pipes (A-C) are widely used in many water systems. Cement linings are also commonly applied in-situ after pipe cleaning, usually to prevent the recurrence of red water or tuberculation problems. Unfortunately, little consideratio...

Air conditioning device for reactor buildings

International Nuclear Information System (INIS)

Kikuchi, Shiro.

1982-01-01

Purpose: To decrease the opening areas of pipe lines for an air conditioning device at the portions passing through the shielding walls of a reactor building for a FBR type reactor, as well as reduce the size of the building. Constitution: Airs in the building for containing reactor are liquefied in an air liquefying mechanism. The liquefied airs are sent by way of pipe lines to each of evaporators, wherein each of the chambers are cooled because of latent heat of evaporation and evaporated airs are released to each of the chambers. The airs released to each of the chambers are collected into an exhaust chamber and sent by way of a duct to the air liquefying mechanism and liquefied again. Since the volume of the liquefied airs may be smaller than the amount conventionally required for usual cooled airs, the pipe lines passing through the shielding walls of the building can be of smaller diameter. This can decrease the opening areas of the pipe lines at the portions passing through the walls of the shieldings and, since the opening areas are smaller, the structure of the radiation shieldings can be simplified in these portions. Further, since the space of the pipe lines in the building is reduced extremely, the size of the building can be reduced. (Moriyama, K.)

Reactor primary coolant system pipe rupture study. Progress report No. 33, January--June 1975

International Nuclear Information System (INIS)

1975-10-01

The pipe rupture study is designed to extend the understanding of failure-causing mechanisms and to provide improved capability for evaluating reactor piping systems to minimize the probability of failures. Following a detailed review to determine the effort most needed to improve nuclear system piping (Phase 1), analytical and experimental efforts (Phase 2) were started in 1965. This progress report summarizes the recent accomplishments of a broad program in (a) basic fatigue crack growth rate studies focused on LWR primary piping materials in a simulated BWR primary coolant environment, (b) at-reactor tests of the effect of primary coolant environment on the fatigue behavior of piping steels, (c) studies directed at quantifying weld sensitization in Type 304 stainless steel, (d) support studies to characterize the electrochemical potential behavior of a typical BWR primary water environment and (e) special tests related to simulation of fracture surfaces characteristic of IGSCC field failures

Pipe connector

International Nuclear Information System (INIS)

Sullivan, T.E.; Pardini, J.A.

1978-01-01

A safety test facility for testing sodium-cooled nuclear reactor components includes a reactor vessel and a heat exchanger submerged in sodium in the tank. The reactor vessel and heat exchanger are connected by an expansion/deflection pipe coupling comprising a pair of coaxially and slidably engaged tubular elements having radially enlarged opposed end portions of which at least a part is of spherical contour adapted to engage conical sockets in the ends of pipes leading out of the reactor vessel and in to the heat exchanger. A spring surrounding the pipe coupling urges the end portions apart and into engagement with the spherical sockets. Since the pipe coupling is submerged in liquid a limited amount of leakage of sodium from the pipe can be tolerated

Critical evaluation of molybdenum and its alloys for use in space reactor core heat pipes

International Nuclear Information System (INIS)

Lundberg, L.B.

1981-01-01

The choice of pure molybdenum as the prime candidate material for space reactor core heat pipes is examined, and the advantages and disadvantages of this material are brought into focus. Even though pure molybdenum heat pipes have been built and tested, this metal's high ductile-brittle transition temperature and modest creep strength place significant design restrictions on a core heat pipe made from it. Molybdenum alloys are examined with regard to their promise as potential replacements for pure molybdenum. The properties of TZM and molybdenum-rhenium alloys are examined, and it appears that Mo-Re alloys with 10 to 15 wt % rhenium offer the most advantage as an alternative to pure molybdenum in space reactor core heat pipes

Assessment of cracked pipes in primary piping systems of PWR nuclear reactors

International Nuclear Information System (INIS)

Jong, Rudolf Peter de

2004-01-01

Pipes related to the Primary System of Pressurized Water Reactors (PWR) are manufactured from high toughness austenitic and low alloy ferritic steels, which are resistant to the unstable growth of defects. A crack in a piping system should cause a leakage in a considerable rate allowing its identification, before its growth could cause a catastrophic rupture of the piping. This is the LBB (Leak Before Break) concept. An essential step in applying the LBB concept consists in the analysis of the stability of a postulated through wall crack in a specific piping system. The methods for the assessment of flawed components fabricated from ductile materials require the use of Elasto-Plastic Fracture Mechanics (EPFM). Considering that the use of numerical methods to apply the concepts of EPFM may be expensive and time consuming, the existence of the so called simplified methods for the assessment of flaws in piping are still considered of great relevance. In this work, some of the simplified methods, normalized procedures and criteria for the assessment of the ductile behavior of flawed components available in literature are described and evaluated. Aspects related to the selection of the material properties necessary for the application of these methods are also discussed. In a next .step, the methods are applied to determine the instability load in some piping configurations under bending and containing circumferential through wall cracks. Geometry and material variations are considered. The instability loads, obtained for these piping as the result of the application of the selected methods, are analyzed and compared among them and with some experimental results obtained from literature. The predictions done with the methods demonstrated that they provide consistent results, with good level of accuracy with regard to the determination of maximum loads. These methods are also applied to a specific Study Case. The obtained results are then analyzed in order to give

76 FR 61682 - Panhandle Eastern Pipe Line Company, LP; Notice of Application

Science.gov (United States)

2011-10-05

... DEPARTMENT OF ENERGY Federal Energy Regulatory Commission [Docket No. CP11-546-000] Panhandle Eastern Pipe Line Company, LP; Notice of Application On September 16, 2011, Panhandle Eastern Pipe Line... free). For TTY, call (202) 502-8659. Comment Date: 5 p.m. Eastern Time on October 19, 2011. Dated...

78 FR 41369 - Certain Small Diameter Carbon and Alloy Seamless Standard, Line and Pressure Pipe From Romania...

Science.gov (United States)

2013-07-10

... Carbon and Alloy Seamless Standard, Line and Pressure Pipe From Romania: Preliminary Results of..., line and pressure pipe (small diameter seamless pipe) from Romania. The period of review (POR) is... and Alloy Seamless Standard, Line and Pressure Pipe from Romania,'' dated concurrently with this...

Specialist meeting on leak before break in reactor piping and vessels

Energy Technology Data Exchange (ETDEWEB)

Bartholome, G.; Bazant, E.; Wellein, R. [Siemens KWU, Stuttgart (Germany)] [and others

1997-04-01

A series of research projects sponsored by the Federal Minister for Education, Science, Research and Technology, Bonn are summarized and compared to utility, manufacturer, and vendor tests. The purpose of the evaluation was to experimentally verify Leak-before-Break behavior, confirm the postulation of fracture preclusion for piping (straight pipe, bends and branches), and quantify the safety margin against massive failure. The results are applicable to safety assessment of ferritic and austenitic piping in primary and secondary nuclear power plant circuits. Moreover, because of the wide range of the test parameters, they are also important for the design and assessment of piping in other technical plant. The test results provide justification for ruling out catastrophic fractures, even on pipes of dimensions corresponding to those of a main coolant pipe of a pressurized water reactor plant on the basis of a mechanical deterministic safety analysis in correspondence with the Basis Safety Concept (Principle of Fracture Exclusion).

Investigation of two-phase flow structure in model of draught pipe of water boiling reactor VK-300

International Nuclear Information System (INIS)

Efanov, A.D.; Kuznetzov, Y.N.; Kaliakin, S.G.; Lisitza, F.D.; Remizov, O.V.; Serdun, N.P.

2001-01-01

VK-300 reactor represents a vessel-type boiling reactor with integral arrangement of assemblies and in-vessel steam separation at one-circuit scheme. The circuit consists of core, draught pipes, and separation facilities. The vessel of VK-300 reactor is chosen on the base of the dimensions of that of VVER-1000 reactor. The following thermal-hydraulic parameters of nuclear power plant (NPP) were investigated experimentally: dependence of void fraction upon the steam quality in mixing chamber (on the draught section input); pressure losses at different, specific zones of up-flow and down-flow sections of the circuit with free circulation; degree of steam separation in the separating chamber (at the first step of phase separation) and its dependence upon steam quality; structure of steam-water flow in draught pipes (distribution of phases over the draught pipe cross- section); presence of steam hovering and height of this hovering in inter-pipe space of draught section. (author)

Challenges associated with the current processes for ultrasonic inspection of CANDU reactor feeder piping

Energy Technology Data Exchange (ETDEWEB)

Machowski, C. [Babcock & Wilcox Canada Ltd., Cambridge, Ontario (Canada)

2012-07-01

CANDU® PHT Feeder Piping is generally constructed from SA106 Grade B carbon steel, which is known to be susceptible to flow accelerated corrosion when exposed to certain environmental conditions. The configuration of the CANDU reactor promotes thinning of the inside surface of the pipe walls, predominantly at the outlet feeders. Inspection of this piping is currently conducted using ultrasonic techniques and is governed by the requirements established by the CANDU Owners Group (COG). There are many challenges associated with these inspections as a result of the complexity of the reactor piping configuration. Geometrical anomalies on the surface of the pipe and non-circular geometries at the tight radius bends hinder the performance of conventional ultrasonic techniques. This can cause lost signals in areas of interest, which in turn often results in rework in order to satisfy the inspection requirements and justify fitness for service of these components. There are also many inspection sites which have limited access due to physical restrictions on the reactor face; therefore in order to maximize the performance of an inspection campaign, it is paramount that the inspection personnel and the inspection technology be well integrated through training simulations prior to execution. These inspection challenges increase the complexity of the analysis process as ultrasonic signals get distorted and lost as a result of non-circular pipe geometries. In order to ensure a high level of integrity in the analysis results, a conservative process is utilized in which two analysts independently examine the data, and a third analyst reviews their results and submits the final call. A Data Management Software application (DMS) is used to input and store the three analysis results. Another important function of the DMS is to provide a communication link between the different work-groups associated with the inspection activities. The focus of this presentation discusses: �

An underground nuclear power station using self-regulating heat-pipe controlled reactors

Science.gov (United States)

Hampel, V.E.

1988-05-17

A nuclear reactor for generating electricity is disposed underground at the bottom of a vertical hole that can be drilled using conventional drilling technology. The primary coolant of the reactor core is the working fluid in a plurality of thermodynamically coupled heat pipes emplaced in the hole between the heat source at the bottom of the hole and heat exchange means near the surface of the earth. Additionally, the primary coolant (consisting of the working fluid in the heat pipes in the reactor core) moderates neutrons and regulates their reactivity, thus keeping the power of the reactor substantially constant. At the end of its useful life, the reactor core may be abandoned in place. Isolation from the atmosphere in case of accident or for abandonment is provided by the operation of explosive closures and mechanical valves emplaced along the hole. This invention combines technology developed and tested for small, highly efficient, space-based nuclear electric power plants with the technology of fast- acting closure mechanisms developed and used for underground testing of nuclear weapons. This invention provides a nuclear power installation which is safe from the worst conceivable reactor accident, namely, the explosion of a nuclear weapon near the ground surface of a nuclear power reactor. 5 figs.

Evaluation of burst pressure prediction models for line pipes

Energy Technology Data Exchange (ETDEWEB)

Zhu, Xian-Kui, E-mail: zhux@battelle.org [Battelle Memorial Institute, 505 King Avenue, Columbus, OH 43201 (United States); Leis, Brian N. [Battelle Memorial Institute, 505 King Avenue, Columbus, OH 43201 (United States)

2012-01-15

Accurate prediction of burst pressure plays a central role in engineering design and integrity assessment of oil and gas pipelines. Theoretical and empirical solutions for such prediction are evaluated in this paper relative to a burst pressure database comprising more than 100 tests covering a variety of pipeline steel grades and pipe sizes. Solutions considered include three based on plasticity theory for the end-capped, thin-walled, defect-free line pipe subjected to internal pressure in terms of the Tresca, von Mises, and ZL (or Zhu-Leis) criteria, one based on a cylindrical instability stress (CIS) concept, and a large group of analytical and empirical models previously evaluated by Law and Bowie (International Journal of Pressure Vessels and Piping, 84, 2007: 487-492). It is found that these models can be categorized into either a Tresca-family or a von Mises-family of solutions, except for those due to Margetson and Zhu-Leis models. The viability of predictions is measured via statistical analyses in terms of a mean error and its standard deviation. Consistent with an independent parallel evaluation using another large database, the Zhu-Leis solution is found best for predicting burst pressure, including consideration of strain hardening effects, while the Tresca strength solutions including Barlow, Maximum shear stress, Turner, and the ASME boiler code provide reasonably good predictions for the class of line-pipe steels with intermediate strain hardening response. - Highlights: Black-Right-Pointing-Pointer This paper evaluates different burst pressure prediction models for line pipes. Black-Right-Pointing-Pointer The existing models are categorized into two major groups of Tresca and von Mises solutions. Black-Right-Pointing-Pointer Prediction quality of each model is assessed statistically using a large full-scale burst test database. Black-Right-Pointing-Pointer The Zhu-Leis solution is identified as the best predictive model.

Evaluation of burst pressure prediction models for line pipes

International Nuclear Information System (INIS)

Zhu, Xian-Kui; Leis, Brian N.

2012-01-01

Accurate prediction of burst pressure plays a central role in engineering design and integrity assessment of oil and gas pipelines. Theoretical and empirical solutions for such prediction are evaluated in this paper relative to a burst pressure database comprising more than 100 tests covering a variety of pipeline steel grades and pipe sizes. Solutions considered include three based on plasticity theory for the end-capped, thin-walled, defect-free line pipe subjected to internal pressure in terms of the Tresca, von Mises, and ZL (or Zhu-Leis) criteria, one based on a cylindrical instability stress (CIS) concept, and a large group of analytical and empirical models previously evaluated by Law and Bowie (International Journal of Pressure Vessels and Piping, 84, 2007: 487–492). It is found that these models can be categorized into either a Tresca-family or a von Mises-family of solutions, except for those due to Margetson and Zhu-Leis models. The viability of predictions is measured via statistical analyses in terms of a mean error and its standard deviation. Consistent with an independent parallel evaluation using another large database, the Zhu-Leis solution is found best for predicting burst pressure, including consideration of strain hardening effects, while the Tresca strength solutions including Barlow, Maximum shear stress, Turner, and the ASME boiler code provide reasonably good predictions for the class of line-pipe steels with intermediate strain hardening response. - Highlights: ► This paper evaluates different burst pressure prediction models for line pipes. ► The existing models are categorized into two major groups of Tresca and von Mises solutions. ► Prediction quality of each model is assessed statistically using a large full-scale burst test database. ► The Zhu-Leis solution is identified as the best predictive model.

Determination of two dimensional axisymmetric finite element model for reactor coolant piping nozzles

International Nuclear Information System (INIS)

Choi, S. N.; Kim, H. N.; Jang, K. S.; Kim, H. J.

2000-01-01

The purpose of this paper is to determine a two dimensional axisymmetric model through a comparative study between a three dimensional and an axisymmetric finite element analysis of the reactor coolant piping nozzle subject to internal pressure. The finite element analysis results show that the stress adopting the axisymmetric model with the radius of equivalent spherical vessel are well agree with that adopting the three dimensional model. The radii of equivalent spherical vessel are 3.5 times and 7.3 times of the radius of the reactor coolant piping for the safety injection nozzle and for the residual heat removal nozzle, respectively

Calculation of forces on reactor containment fan cooler piping

International Nuclear Information System (INIS)

Miller, J.S.; Ramsden, K.

2004-01-01

The purpose of this paper is to present the results of the Reactor Containment Fan Cooler (RCFC) system piping load calculations. These calculations are based on piping loads calculated using the EPRI methodology and RELAP5 to simulate the hydraulic behavior of the system. The RELAP5 generated loads were compared to loads calculated using the EPRI GL-96-06 methodology. This evaluation was based on a pressurized water reactor's RCFC coils thermal hydraulic behavior during a Loss of Offsite Power (LOOP) and a loss of coolant accident (LOCA). The RCFC consist of two banks of service water and chill water coils. There are 5 SX and 5 chill water coils per bank. Therefore, there are 4 RCFC units in the containment with 2 banks of coils per RCFC. Two Service water pumps provide coolant for the 4 RCFC units (8 banks total, 2 banks per RCFC unit and 2 RCFC units per pump). Following a LOOP/LOCA condition, the RCFC fans would coast down and upon being re-energized, would shift to low-speed operation. The fan coast down is anticipated to occur very rapidly due to the closure of the exhaust damper as a result of LOCA pressurization effects. The service water flow would also coast down and be restarted in approximately 43 seconds after the initiation of the event. The service water would drain from the RCFC coils during the pump shutdown and once the pumps restart, water is quickly forced into the RCFC coils causing hydraulic loading on the piping. Because of this scenario and the potential for over stressing the piping, an evaluation was performed by the utility using RELAP5 to assess the piping loads. Subsequent to the hydraulic loads being analyzed using RELAP5, EPRI through GL-96-06 provided another methodology to assess loads on the RCFC piping system. This paper presents the results of using the EPRI methodology and RELAP5 to perform thermal hydraulic load calculations. It is shown that both EPRI methodology and RELAP5 calculations can be used to generate hydraulic loads

Water inlet and steam outlet pipes fitted one inside the other for nuclear reactors

International Nuclear Information System (INIS)

Mc Donald, B.N.

1976-01-01

A description is given of a combined exhaust nozzle and intake pipe system to support a heat exchanger inside a nuclear reactor pressure vessel. It comprises a generally cylindrical part on the exhaust nozzle, the cylindrical part having an inside passage, a flange around the passage and provided with means to secure the exhaust nozzle to the reactor pressure vessel so as to make it fluidtight. The cylindrical part has an aperture inside to take the intake pipe inside the passage so as to enable the intake pipe to project into the heat exchanger. A collar made on the heat exchanger projects from the heat exchanger to the cylindrical nozzle component to establish communication with the inside passage for the fluid [fr

Hydrogen permeation resistant heat pipe for bi-modal reactors. Final report, October 1, 1994--September 30, 1995

International Nuclear Information System (INIS)

North, M.T.; Anderson, W.G.

1995-01-01

The principal objective of this program was to demonstrate technology that will make a sodium heat pipe tolerant of hydrogen permeation for a bimodal space reactor application. Special focus was placed on techniques which enhance the permeation of hydrogen out of the heat pipe. Specific objectives include: define the detailed requirements for the bimodal reactor application; design and fabricate a prototype heat pipe tolerant of hydrogen permeation; and test the prototype heat pipe and demonstrate that hydrogen which permeates into the heat pipe is removed or reduced to acceptable levels. The results of the program were fully successful. Analyses were performed on two different heat pipe designs and an experimental heat pipe was fabricated and tested. A model of the experimental heat pipe was developed to predict the enhancement in the hydrogen permeation rate out of the heat pipe. A significant improvement in the rate at which hydrogen permeates out of a heat pipe was predicted for the use of the special condenser geometry developed here. Agreement between the model and the experimental results was qualitatively good. Inclusion of the additional effects of fluid flow in the heat pipe are recommended for future work

Detection system for location of fuel pebbles transported in pipes in a pebble-bed reactor based on vibration signal processing

Energy Technology Data Exchange (ETDEWEB)

Liu, Hongbing, E-mail: liuhb07@mails.tsinghua.edu.cn [Department of Mechanical Engineering, Tsinghua University, Beijing 100084 (China); Key Laboratory for Advanced Materials Processing Technology, Ministry of Education P. R. China, Beijing 100084 (China); Du, Dong, E-mail: dudong@tsinghua.edu.cn [Department of Mechanical Engineering, Tsinghua University, Beijing 100084 (China); Key Laboratory for Advanced Materials Processing Technology, Ministry of Education P. R. China, Beijing 100084 (China); Huang, An; Chang, Baohua; Han, Zandong [Department of Mechanical Engineering, Tsinghua University, Beijing 100084 (China); Key Laboratory for Advanced Materials Processing Technology, Ministry of Education P. R. China, Beijing 100084 (China); He, Ayada [Shanghai Electric Power Generation Group Shanghai Generator Works, Shanghai 200240 (China)

2016-08-15

Highlights: • A detection system for locations of pebbles transported in pipes is introduced. • The detection system is based on vibration signal processing, which is original. • The characteristics of the vibration signals of the pipe are analyzed. • The experiment shows that the detection results are accurate. • The research provides an important basis for the design of the reactor. - Abstract: Pebble-bed high temperature gas-cooled reactors have many advantages such as inherent safety, high efficiency, etc., and have been considered as a candidate for Generation IV nuclear reactors. During the operation of the reactor, there are thousands of fuel pebbles transported in the pipes outside the core by gravity and helium flow. The pattern of the pipes which consist of straight and arc sections is very complex. When a fuel pebble is transported, it will constantly collide with the pipes, especially in the arc sections. The collisions will lead to the vibration of the pipes. This paper aims to provide a detection system for the location of fuel pebbles transported in pipes in a pebble-bed reactor based on vibration signal processing. Before the reactor is running, the system acquires the vibration signals of several key sections by sensors. Then the frequency characteristics of the signals are obtained by joint time–frequency analysis. When the reactor is running, the system detects the signals and processes them based on their frequency characteristics in real time. According to the results of the processing, the system can correctly judge whether the fuel pebble has passed through the section and records the time of the passing. The experiment validates the accuracy and reliability of the detection results. In this way, the operational condition of the reactor can be monitored so that the normal running of the reactor can be ensured. Additionally, the detection data are of great significance to the evaluation and optimization of the reactor performance

Detection system for location of fuel pebbles transported in pipes in a pebble-bed reactor based on vibration signal processing

International Nuclear Information System (INIS)

Liu, Hongbing; Du, Dong; Huang, An; Chang, Baohua; Han, Zandong; He, Ayada

2016-01-01

Highlights: • A detection system for locations of pebbles transported in pipes is introduced. • The detection system is based on vibration signal processing, which is original. • The characteristics of the vibration signals of the pipe are analyzed. • The experiment shows that the detection results are accurate. • The research provides an important basis for the design of the reactor. - Abstract: Pebble-bed high temperature gas-cooled reactors have many advantages such as inherent safety, high efficiency, etc., and have been considered as a candidate for Generation IV nuclear reactors. During the operation of the reactor, there are thousands of fuel pebbles transported in the pipes outside the core by gravity and helium flow. The pattern of the pipes which consist of straight and arc sections is very complex. When a fuel pebble is transported, it will constantly collide with the pipes, especially in the arc sections. The collisions will lead to the vibration of the pipes. This paper aims to provide a detection system for the location of fuel pebbles transported in pipes in a pebble-bed reactor based on vibration signal processing. Before the reactor is running, the system acquires the vibration signals of several key sections by sensors. Then the frequency characteristics of the signals are obtained by joint time–frequency analysis. When the reactor is running, the system detects the signals and processes them based on their frequency characteristics in real time. According to the results of the processing, the system can correctly judge whether the fuel pebble has passed through the section and records the time of the passing. The experiment validates the accuracy and reliability of the detection results. In this way, the operational condition of the reactor can be monitored so that the normal running of the reactor can be ensured. Additionally, the detection data are of great significance to the evaluation and optimization of the reactor performance

Fatigue check of nuclear safety class 1 reactor coolant pipe

International Nuclear Information System (INIS)

Wang Qing; Fang Yonggang; Chu Qibao; Xu Yu; Li Hailong

2015-01-01

Fatigue and thermal ratcheting analyses of nuclear safety Class 1 reactor coolant pipe in a nuclear power plant were independently carried out in this paper. The software used for calculation is ROCOCO, which is based on RCC-M code. The difference of nuclear safety Class 1 pipe fatigue evaluation between RCC-M code and ASME code was compared. The main aspects of comparison include the calculation scoping of fatigue design, the calculation method of primary plus secondary stress intensity, the elastic-plastic correction coefficient calculation, and the dynamic load combination method etc. By correcting inconsistent algorithm of ASME code within ROCOCO, the fatigue usage factor and thermal ratcheting design margin of 65 mm and 55 mm wall thickness of the pipe were obtained. The results show that the minimum wall thickness of the pipe must exceed 55 mm and the design value of the thermal ratcheting of 55 mm wall thickness reaches 95% of the allowable value. (authors)

Screening reactor steam/water piping systems for water hammer

International Nuclear Information System (INIS)

Griffith, P.

1997-09-01

A steam/water system possessing a certain combination of thermal, hydraulic and operational states, can, in certain geometries, lead to a steam bubble collapse induced water hammer. These states, operations, and geometries are identified. A procedure that can be used for identifying whether an unbuilt reactor system is prone to water hammer is proposed. For the most common water hammer, steam bubble collapse induced water hammer, six conditions must be met in order for one to occur. These are: (1) the pipe must be almost horizontal; (2) the subcooling must be greater than 20 C; (3) the L/D must be greater than 24; (4) the velocity must be low enough so that the pipe does not run full, i.e., the Froude number must be less than one; (5) there should be void nearby; (6) the pressure must be high enough so that significant damage occurs, that is the pressure should be above 10 atmospheres. Recommendations on how to avoid this kind of water hammer in both the design and the operation of the reactor system are made

Investigation and examination on the cracking of pipings in boiling water reactors

International Nuclear Information System (INIS)

1977-01-01

This is the report made by the Reactor Safety Technology Expert Committee to the Atomic Energy Commission regarding the investigation and examination on stress corrosion cracking which seems to be the cause of the cracking of pipings in boiling water reactors, the measures to reduce it, and the subjects of research hereafter. Recently, the stress corrosion cracking of primary coolant pipings has been often observed, and this phenomenon occurred in the pressure boundary of primary coolant, consequently it is possible to be linked to the troubles of large scale. The Reactor Material Subcommittee was established on May 14, 1975, and investigated the cracking phenomena in the recirculating system and core spray system of BWRs in Japan and foreign countries. The recent cases have been concentrated to the heat-affected part due to welding of 304 type austenitic stainless steel pipings of from 4 in to 10 in diameter for BWRs. They are the stress corrosion cracking at grain boundaries occurred under the loaded condition and in the environment of high temperature, high pressure water. The cracking of this kind was never experienced in PWRs. The results of the technical examination, the consideration of the mechanism of stress corrosion cracking, and the countermeasures are described. (Kako, I.)

77 FR 67336 - Certain Small Diameter Carbon and Alloy Seamless Standard, Line and Pressure Pipe From Romania...

Science.gov (United States)

2012-11-09

... Carbon and Alloy Seamless Standard, Line and Pressure Pipe From Romania: Final Results of Antidumping... alloy seamless standard, line and pressure pipe from Romania. The period of review is August 1, 2010..., line and pressure pipe from Romania. See Certain Small Diameter Carbon and Alloy Seamless Standard...

Application of ultrasonic testing technique to detect gas accumulation in important pipings for pressurized water reactors safety

Energy Technology Data Exchange (ETDEWEB)

Fushimi, Yasuyuki [Inst. of Nuclear Safety System Inc., Mihama, Fukui (Japan)

2002-09-01

Since 1988, the USNRC has pointed out that gas-binding events might occur at high head safety injection (HHSI) pumps of pressurized water reactors (PWRs). In Japanese PWR plants, corrective actions were taken in response to gas-binding events that occurred on HHSI pumps in the USA, so no gas accumulation event has been reported so far. However, when venting frequency is prolonged with operating cycle extension, the probability of gas accumulation in pipings may increase as in the USA. The purpose of this study was to establish a technique to identify gas accumulation and to measure the gas volume accurately. Taking dominant causes of the gas-binding events in the USA into consideration, we pointed out the following sections in the Japanese PWRs where gas srtipping and/or gas accumulation might occur: residual heat removal system pipings and charging/safety injection pump minimum flow line. Then an ultrasonic testing technique, adopted to identify gas accumulation in the USA, was applied to those sections of the typical Japanese PWR. Consequently, no gas accumulation was found in those pipings. (author)

76 FR 60083 - Carbon and Alloy Seamless Standard, Line, and Pressure Pipe From Japan and Romania

Science.gov (United States)

2011-09-28

... Alloy Seamless Standard, Line, and Pressure Pipe From Japan and Romania Determinations On the basis of... pressure pipe from Japan and Romania would be likely to lead to continuation or recurrence of material... regarding small- diameter carbon and alloy seamless standard, line, and pressure pipe from Romania...

Experimental and analytical studies on creep failure of reactor coolant piping

Energy Technology Data Exchange (ETDEWEB)

Maeda, Akio; Maruyama, Yu; Hashimoto, Kazuichiro; Harada, Yuhei; Shibazaki, Hiroaki; Kudo, Tamotsu; Hidaka, Akihide; Sugimoto, Jun [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment; Nakamura, N.

1999-07-01

Thermal and structural responses of reactor coolant piping under and elevated internal pressure and temperature are being investigated in WIND project at JAERI. In a recent failure test in which a nuclear grade type 316 stainless steel pipe with an outer diameter of 114.3 mm and a wall thickness of 13.5 mm was used and an internal pressure was kept at approximately 15 MPa. A failure of the piping was observed when the temperature was sustained at 970degC for one hour. In parallel with conducting the tests, post-test analyses were performed. The objective of the analyses is to assess analytical models for the creep deformation and failure of the piping at elevated internal pressure and temperature simulating thermal-hydraulic conditions during a severe accident. The major material properties needed for the analysis were measured at elevated temperatures. Coefficients of a creep constitutive equation including the tertiary stage were determined with the measured creep data and incorporated into ABAQUS code. The analysis reasonably reproduced the time history of the enlargement of the piping diameter, and the wall thickness and the diameter of the piping at the failure. It was also found that the piping failure timing obtained from the analysis agreed well with the test result. (author)

Experimental and analytical studies on creep failure of reactor coolant piping

International Nuclear Information System (INIS)

Maeda, Akio; Maruyama, Yu; Hashimoto, Kazuichiro; Harada, Yuhei; Shibazaki, Hiroaki; Kudo, Tamotsu; Hidaka, Akihide; Sugimoto, Jun; Nakamura, N.

1999-01-01

Thermal and structural responses of reactor coolant piping under and elevated internal pressure and temperature are being investigated in WIND project at JAERI. In a recent failure test in which a nuclear grade type 316 stainless steel pipe with an outer diameter of 114.3 mm and a wall thickness of 13.5 mm was used and an internal pressure was kept at approximately 15 MPa. A failure of the piping was observed when the temperature was sustained at 970degC for one hour. In parallel with conducting the tests, post-test analyses were performed. The objective of the analyses is to assess analytical models for the creep deformation and failure of the piping at elevated internal pressure and temperature simulating thermal-hydraulic conditions during a severe accident. The major material properties needed for the analysis were measured at elevated temperatures. Coefficients of a creep constitutive equation including the tertiary stage were determined with the measured creep data and incorporated into ABAQUS code. The analysis reasonably reproduced the time history of the enlargement of the piping diameter, and the wall thickness and the diameter of the piping at the failure. It was also found that the piping failure timing obtained from the analysis agreed well with the test result. (author)

Primary coolant pipe rupture event in liquid metal cooled reactors. Proceedings of a technical meeting

International Nuclear Information System (INIS)

2004-08-01

In liquid-metal cooled fast reactors (LMFR) the primary coolant pipes (PCP) connect the primary coolant pumps to the grid plate. A rupture in one of these pipes could cause significant loss of coolant flow to the core with severe consequences. In loop type reactors, all primary pipelines are provided with double envelopes and inter-space coolant leak monitoring systems that permit leak detection before break. Thus, the PCP rupture event can be placed in the beyond design basis event (BDBE) category. Such an arrangement is difficult to incorporate for pool type reactors, and hence it could be argued that the PCP rupture event needs to be analysed in detail as a design basis event (DBE, category 4 event). However, the primary coolant pipes are made of ductile austenitic stainless steel material and operate at temperatures of the cold pool and at comparatively low pressures. For such low stressed piping with negligible creep and embrittlement effects, it is of interest to discuss under what design provisions, for pool type reactors, the guillotine rupture of PCP could be placed in the BDBE category. The topical Technical Meeting (TM) on Primary Coolant Pipe Rupture Event in Liquid Metal Cooled Reactors (Indira Gandhi Centre for Atomic Research, Kalpakkam, India, 13-17 January 2003) was called to enable the specialists to present the philosophy and analyses applied on this topic in the various Member States for different LMFRs. The scope of the technical meeting was to provide a global forum for information exchange on the philosophy applied in the various participating Member States and the analyses performed for different LMFRs with regard to the primary coolant pipe rupture event. More specifically, the objectives of the technical meeting were to review the safety philosophy for the PCP rupture event in pool type LMFR, to assess the structural reliability of the PCP and the probability of rupture under different conditions (with/without in-service inspection), to

Sodium heat pipe module test for the SAFE-30 reactor prototype

International Nuclear Information System (INIS)

Reid, Robert S.; Sena, J. Tom; Martinez, Adam L.

2001-01-01

Reliable, long-life, low-cost heat pipes can enable safe, affordable space fission power and propulsion systems. Advanced versions of these systems can in turn allow rapid access to any point in the solar system. Twelve stainless steel-sodium heat pipe modules were built and tested at Los Alamos for use in a non-nuclear thermohydraulic simulation of the SAFE-30 reactor (Poston et al., 2000). SAFE-30 is a near-term, low-cost space fission system demonstration. The heat pipes were designed to remove thermal power from the SAFE-30 core, and transfer this power to an electrical power conversion system. These heat pipe modules were delivered to NASA Marshall Space Flight Center in August 2000 and were assembled and tested in a prototypical configuration during September and October 2000. The construction and test of one of the SAFE-30 modules is described

Method for operating nuclear reactor

International Nuclear Information System (INIS)

Utamura, Motoaki; Urata, Megumu; Uchida, Shunsuke

1978-01-01

Purpose: In order to judge the fuel failures, if any, without opening a reactor container for BWR type reactors, a method has been described for measuring the difference between the temperature dependent iodine spike value and the pressure dependent iodine spike value in the pressure vessel. Method: After the scram of a nuclear reactor, steam generated by decay heat is condensed in a remaining heat exchanger and cooling water is returned through a recycling pipe line to a reactor core. At the same time, a control rod drive system pump is operated, the reactor core is filled with the cooling water. Then, the coolant is taken from the recycling pipe line to cool the reactor core. After applying the temperature fluctuation, the cooling water is sampled at a predetermined time interval at a sampling point to determine the changes with time in the radioactive concentration of iodine. When the radioactivity of iodine in the cooling water is lowered sufficiently by a reactor purifying system, the nuclear reactor vessel is depressurized. After applying pressure fluctuation, iodine spike value is determined. (Kawakami, Y.)

Mechanical Properties of Post Irradiation Primary Cooling Piping of Bandung Research Reactor

International Nuclear Information System (INIS)

Histori; Renaningsih S; Sri Nitiswati; Ari Triyadi

2003-01-01

Testing on primary coolant piping of research reactor Bandung have been done. Primary coolant piping were made from Al 6061-T6. The goal of this activity is to investigate the mechanical properties changes caused by aging process after 33 years in irradiated. Type of testing i.e visual examination, thickness measurement, tensile and hardness test were done. The test data shown that there was a deposit at the inside surface of pipe, thickness decreased about 0.2 mm, tensile strength is 293 MPa, yield strength is 262 MPa, while the hardness is about 83 HRE (mean value). The test data than compared with ASTM standard. As the conclusion tensile and yield strength of pipe still fulfill the ASTM requirements, except the hardness is unsignificantly less/decreased. (author)

Innovative technology summary report: High-speed clamshell pipe cutter

International Nuclear Information System (INIS)

1998-09-01

The Hanford Site C Reactor Technology Demonstration Group demonstrated the High-Speed Clamshell Pipe Cutter technology, developed and marketed by Tri Tool Inc. (Rancho Cordova, California). The models demonstrated are portable, split-frame pipe lathes that require minimal radial and axial clearances for severing and/or beveling in-line pipe with ranges of 25 cm to 41 cm and 46 cm to 61 cm nominal diameter. The radial clearance requirement from the walls, floors, or adjacent pipes is 18 cm. The lathes were supplied with carbide insert conversion kits for the cutting bits for the high-speed technique that was demonstrated. Given site-specific factors, this demonstration showed the cost of the improved technology to be approximately 30% higher than the traditional (baseline) technology (oxyacetylene torch) cost of $14,400 for 10 cuts of contaminated 41-cm and 61-cm-diameter pipe at C Reactor. Actual cutting times were faster than the baseline technology; however, moving/staging the equipment took longer. Unlike the baseline torch, clamshell lathes do not involve applied heat, flames, or smoke and can be operated remotely, thereby helping personal exposures to be as low as reasonably achievable. The baseline technology was demonstrated at the C Reactor north and south water pipe tunnels August 19--22, 1997. The improved technology was demonstrated in the gas pipe tunnel December 15--19

75 FR 69125 - Certain Seamless Carbon and Alloy Steel Standard, Line, and Pressure Pipe From China

Science.gov (United States)

2010-11-10

... with material injury by reason of imports from China of certain seamless carbon and alloy steel standard, line, and pressure pipe (``seamless SLP pipe''), provided for in subheadings 7304.19.10, 7304.19... Seamless Carbon and Alloy Steel Standard, Line, and Pressure Pipe From China Determination On the basis of...

Integrity of the reactor coolant boundary of the European pressurized water reactor (EPR)

Energy Technology Data Exchange (ETDEWEB)

Goetsch, D.; Bieniussa, K.; Schulz, H.; Jalouneix, J.

1997-04-01

This paper is an abstract of the work performed in the frame of the development of the IPSN/GRS approach in view of the EPR conceptual safety features. EPR is a pressurized water reactor which will be based on the experience gained by utilities and designers in France and in Germany. The reactor coolant boundary of a PWR includes the reactor pressure vessel (RPV), those parts of the steam generators (SGs) which contain primary coolant, the pressurizer (PSR), the reactor coolant pumps (RCPs), the main coolant lines (MCLs) with their branches as well as the other connecting pipes and all branching pipes including the second isolation valves. The present work covering the integrity of the reactor coolant boundary is mainly restricted to the integrity of the main coolant lines (MCLs) and reflects the design requirements for the main components of the reactor coolant boundary. In the following the conceptual aspects, i.e. design, manufacture, construction and operation, will be assessed. A main aspect is the definition of break postulates regarding overall safety implications.

Integrity of the reactor coolant boundary of the European pressurized water reactor (EPR)

International Nuclear Information System (INIS)

Goetsch, D.; Bieniussa, K.; Schulz, H.; Jalouneix, J.

1997-01-01

This paper is an abstract of the work performed in the frame of the development of the IPSN/GRS approach in view of the EPR conceptual safety features. EPR is a pressurized water reactor which will be based on the experience gained by utilities and designers in France and in Germany. The reactor coolant boundary of a PWR includes the reactor pressure vessel (RPV), those parts of the steam generators (SGs) which contain primary coolant, the pressurizer (PSR), the reactor coolant pumps (RCPs), the main coolant lines (MCLs) with their branches as well as the other connecting pipes and all branching pipes including the second isolation valves. The present work covering the integrity of the reactor coolant boundary is mainly restricted to the integrity of the main coolant lines (MCLs) and reflects the design requirements for the main components of the reactor coolant boundary. In the following the conceptual aspects, i.e. design, manufacture, construction and operation, will be assessed. A main aspect is the definition of break postulates regarding overall safety implications

78 FR 11639 - Houston Pipe Line Company LP; Notice of Petition for Rate Approval

Science.gov (United States)

2013-02-19

... DEPARTMENT OF ENERGY Federal Energy Regulatory Commission [Docket No. PR13-31-000] Houston Pipe Line Company LP; Notice of Petition for Rate Approval Take notice that on February 1, 2013, Houston Pipe Line Company LP (HPL) filed for approval of rates for transportation service pursuant to section...

Proven GIS adaptions by other industries benefit pipe lines

International Nuclear Information System (INIS)

Sanders, M.D.

1994-01-01

Automated mapping (AM) and facilities management (FM) projects in the pipe line industry are becoming increasingly desirable applications of geographic information system (GIS) technology. In the vernacular of GIS technology, application for the pipe line industry are commonly referred to as automated mapping and facilities management (AM/FM). Computer software allows the use of computer aided drafting (CAD) and database packages for information storage/retrieval to provide displays and reports of data set relationships in a given location and area. Geographic information management technology has grown to meet the expanding database resource capabilities. This technological growth also has combined with increasingly powerful and efficient computer systems and networks with plummeting hardware, software, and network enhancement costs. This paper discusses Regulatory compliance, planning, implementation, and data sharing process of the GIS

Vibration analysis of primary inlet pipe line during steady state and transient conditions of Pakistan research reactor-1

International Nuclear Information System (INIS)

Ayazuddin, S.K.; Qureshi, A.A.; Hayat, T.

1997-11-01

The Primary Water Inlet Pipeline (PW-IPL) is of stainless steel conveying demineralized water from hold-up tank to the reactor pool of Pakistan Research Reactor-1 (PARR-1). The section of the pipeline from heat exchangers to the valve pit is hanger supported in the pump room and the rest of the section from valve pit to the reactor pool is embedded. The PW-IPL is subjected to steady state and transient vibrations. The reactor pumps, which drive the coolant through various circuits mainly contribute the steady state vibrations, while transient vibrations arise due to instant closure of the check valve (water hammer). The ASME Boiler and Pressure Vessel code provides data about the acceptable limits of stresses related to the primary static stress due to steady state vibrations. However, due to complexity in the pipe structure, stresses related to the transient vibrations are neglected in the code. In this report attempt has been made to analyzed both steady state and transient vibrations of PW-IPL of PARR-1. Since, both the steady state and transient vibrations affect the hanger-supported section of the PW-IPL, therefore, it was selected for vibration test measurements. In the analysis vibration data was compared with the allowable limits and estimations of maximum pressure build-up, eflection, natural frequency, tensile and shear load on hanger support, and the ratio of maximum combine stress to the allowable load were made. (author)

Transient freezing of molten salts in pipe-flow systems: Application to the direct reactor auxiliary cooling system (DRACS)

International Nuclear Information System (INIS)

Le Brun, N.; Hewitt, G.F.; Markides, C.N.

2017-01-01

Highlights: • A thermo-hydraulic model has been proposed to simulate the transient freezing of molten salts in complex piping systems. • The passive safety system DRACS in Generation-IV, molten salt reactor is susceptible to failure due to salt freezing. • For the prototypical 0.2 MW reactor considered in this study considerable freezing occurs after 20 minutes leading to reactor temperatures above 900 °C within 4 hours. • Conservative criteria for the most important/least known variables in the design of DRACS have been discussed. • Over-conservative approaches in designing the NDHX should be used with caution as they can promote pipe clogging due to freezing. - Abstract: The possibility of molten-salt freezing in pipe-flow systems is a key concern for the solar-energy industry and a safety issue in the new generation of molten-salt reactors, worthy of careful consideration. This paper tackles the problem of coolant solidification in complex pipe networks by developing a transient thermohydraulic model and applying it to the ‘Direct Reactor Auxiliary Cooling System’ (DRACS), the passive-safety system proposed for the Generation-IV molten-salt reactors. The results indicate that DRACS, as currently envisioned, is prone to failure due to freezing in the air/molten-salt heat exchanger, which can occur after approximately 20 minutes, leading to reactor temperatures above 900 °C within 4 hours. The occurrence of this scenario is related to an unstable behaviour mode of DRACS in which newly formed solid-salt deposit on the pipe walls acts to decrease the flow-rate in the secondary loop, facilitating additional solid-salt deposition. Conservative criteria are suggested to facilitate preliminary assessments of early-stage DRACS designs. The present study is, to the knowledge of the authors, the first of its kind in serving to illustrate possible safety concerns in molten-salt reactors, which are otherwise considered very safe in the literature. Furthermore

Insulated pipe clamp design

International Nuclear Information System (INIS)

Anderson, M.J.; Hyde, L.L.; Wagner, S.E.; Severud, L.K.

1980-01-01

Thin wall large diameter piping for breeder reactor plants can be subjected to significant thermal shocks during reactor scrams and other upset events. On the Fast Flux Test Facility, the addition of thick clamps directly on the piping was undesired because the differential metal temperatures between the pipe wall and the clamp could have significantly reduced the pipe thermal fatigue life cycle capabilities. Accordingly, an insulated pipe clamp design concept was developed. 5 refs

Heat transfer characteristics and operation limit of pressurized hybrid heat pipe for small modular reactors

International Nuclear Information System (INIS)

Kim, Kyung Mo; Bang, In Cheol

2017-01-01

Highlights: • Thermal performances and operation limits of hybrid heat pipe were experimentally studied. • Models for predicting the operation limit of the hybrid heat pipe was developed. • Non-condensable gas affected heat transfer characteristics of the hybrid heat pipe. - Abstract: In this paper, a hybrid heat pipe is proposed for use in advanced nuclear power plants as a passive heat transfer device. The hybrid heat pipe combines the functions of a heat pipe and a control rod to simultaneously remove the decay heat generated from the core and shutdown the reactor under accident conditions. Thus, the hybrid heat pipe contains a neutron absorber in the evaporator section, which corresponds to the core of the reactor pressure vessel. The presence of the neutron absorber material leads to differences in the heated diameter and hydraulic diameter of the heat pipe. The cross-sectional areas of the vapor paths through the evaporator, adiabatic, and condenser sections are also different. The hybrid heat pipe must operate in a high-temperature, high-pressure environment to remove the decay heat. In other words, the operating pressure must be higher than those of the commercially available thermosyphons. Hence, the thermal performances, including operation limit of the hybrid heat pipe, were experimentally studied in the operating pressure range of 0.2–20 bar. The operating pressure of the hybrid heat pipe was controlled by charging the non-condensable gas which is unused method to achieve the high saturation pressure in conventional thermosyphons. The effect of operating pressure on evaporation heat transfer was negligible, while condensation heat transfer was affected by the amount of non-condensable gas in the test section. The operation limit of the hybrid heat pipe increased with the operating pressure. Maximum heat removal capacity of the hybrid heat pipe was up to 6 kW which is meaningful value as a passive decay heat removal device in the nuclear power

The feasibility of remotely separating and rejoining the main coolant pipes of a fusion reactor

International Nuclear Information System (INIS)

Briaris, D.A.; Stanbridge, J.R.

1977-09-01

The generic requirement of a fusion reactor that the first wall and other high neutron dose structures be periodically replaced gives rise to a number of complex engineering operations which need to be performed remotely and with a high degree of reliability. Techniques for the remote separation and rejoining of the helium coolant pipes on the Culham Conceptual Tokamak Reactor Mk. II have been investigated in the form of cutting and welding schemes and the use of a mechanical coupling. A mechanical coupling is the more attractive because the reduced complexity of the operations to separate and join the pipes potentially shortens the reactor down-time. Some assessment of remote joint examination and recovery from faults has also been made. (author)

78 FR 63164 - Certain Small Diameter Carbon and Alloy Seamless Standard, Line and Pressure Pipe From Romania...

Science.gov (United States)

2013-10-23

... Carbon and Alloy Seamless Standard, Line and Pressure Pipe From Romania: Final Results of Antidumping... carbon and alloy seamless standard, line and pressure pipe from Romania. For the final results we... pressure pipe from Romania.\\1\\ We invited interested parties to comment on the Preliminary Results. We...

Evaluation of stress histories of reactor coolant loop piping for pipe rupture prediction

International Nuclear Information System (INIS)

Lu, S.C.; Larder, R.A.; Ma, S.M.

1981-01-01

This paper describes the analyses used to evaluate stress histories in the primary coolant loop piping of a selected four-loop PNR power station. In order to make the simulation as realistic as possible, best estimates rather than conservative assumptions were considered throughout. The best estimate solution, however, was aided by a sensitivity study to assess the possible variation of outcomes resulted from uncertainties associated with these assumptions. Sources of stresses considered in the evaluation were pressure, dead weight, thermal expansion, thermal gradients through the pipe wall, residual welding, pump vibrations, and finally seismic excitations. The best estimates of pressure and thermal transient histories arising from plant operations were based on actual plant operation records supplemented by specified plant design conditions. Seismic motions were generated from response spectrum curves developed specifically for the region surrounding the plant site. Stresses due to dead weight and thermal expansion were computed from a three dimensional finite element model which used a combination of pipe, truss, and beam elements to represent the coolant loop piping, the pressure vessel, coolant pumps, steam generators, and the pressurizer. Stresses due to pressure and thermal gradients were obtained by closed form solutions. Seismic stress calculations considered the soil structure interaction, the coupling effect between the containment structure and the reactor coolant system. A time history method was employed for the seismic analysis. Calculations of residual stresses accounted for the actual heat impact, welding speed, weld preparation geometry, and pre- and post-heat treatments. Vibrational stresses due to pump operation were estimated by a dynamic analysis using existing measurements of pump vibrations. (orig./HP)

Single failure effects of reactor coolant system large bore hydraulic snubbers for Korean Standard Nuclear Power Plant

International Nuclear Information System (INIS)

Choi, T.S.; Park, S.H.; Sung, K.K.; Kim, T.W.; Jheon, J.H.

1996-01-01

A potential snubber single failure is one of the safety significances identified in General Safety Issue 113 for Large Bore Hydraulic Snubber (LBHS) dynamic qualification. This paper investigates dynamic structural effects of single failures of the steam generator and reactor coolant pump snubbers in Korean Standard Nuclear Power Plant by performing the time history dynamic analyses for the reactor coolant system under seismic and postulated pipe break events. The seismic input motions considered are the synthesized ground time histories conforming to SRP 3.7.1, and he postulated pipe break input loadings result from steam generator main seam line and feedwater line pipe breaks which govern pipe breaks remaining after applying LBB to the main coolant line and primary side ranch lines equal to and greater than 12 inch nominal pipe size

Heat Pipe Reactor Dynamic Response Tests: SAFE-100 Reactor Core Prototype

Science.gov (United States)

Bragg-Sitton, Shannon M.

2005-01-01

The SAFE-I00a test article at the NASA Marshall Space Flight Center was used to simulate a variety of potential reactor transients; the SAFEl00a is a resistively heated, stainless-steel heat-pipe (HP)-reactor core segment, coupled to a gas-flow heat exchanger (HX). For these transients the core power was controlled by a point kinetics model with reactivity feedback based on core average temperature; the neutron generation time and the temperature feedback coefficient are provided as model inputs. This type of non-nuclear test is expected to provide reasonable approximation of reactor transient behavior because reactivity feedback is very simple in a compact fast reactor (simple, negative, and relatively monotonic temperature feedback, caused mostly by thermal expansion) and calculations show there are no significant reactivity effects associated with fluid in the HP (the worth of the entire inventory of Na in the core is .tests, the point kinetics model was based on core thermal expansion via deflection measurements. It was found that core deflection was a strung function of how the SAFE-100 modules were fabricated and assembled (in terms of straightness, gaps, and other tolerances). To remove the added variable of how this particular core expands as compared to a different concept, it was decided to use a temperature based feedback model (based on several thermocouples placed throughout the core).

Technical meeting on 'Primary coolant pipe rupture event in liquid metal cooled fast reactors'. Working material

International Nuclear Information System (INIS)

2003-01-01

In Liquid Metal cooled Fast Reactors (LMFR) or in accelerator driven sub-critical systems (ADS) with LMFR like sub-critical cores, the primary coolant pipes (PCP) connect the primary coolant pumps to the grid plate. A rupture in one of these pipes could cause significant loss of coolant flow to the core with severe consequences. In loop type reactors, all primary pipelines are provided with double envelopes and inter-space coolant leak monitoring systems that permit leak detection before break. Thus, the PCP rupture event can be placed in the beyond design basis event (BDBE) category. Such an arrangement is difficult to incorporate for pool type reactors, and hence it could be argued that the PCP rupture event needs to be analysed in detail as a design basis event (DBE, category 4 event). The primary coolant pipes are made of ductile austenitic stainless steel material and operate at temperatures of the cold pool and at comparatively low pressures. For such low stressed piping with negligible creep and embrittlement effects, it is of interest to discuss under what design provisions, for pool type reactors, the guillotine rupture of PCP could be placed in the BDBE category. The topical Technical Meeting (TM) on 'Primary Coolant Pipe Rupture Event in Liquid Metal Cooled Reactors' was called to enable the specialists to present the philosophy and analyses applied on this topic in the various Member States for different LMFRs. The scope of the Technical Meeting was to provide a global forum for information exchange on the philosophy applied in the various participating Member States and the analyses performed for different LMFRs with regard to the primary coolant pipe rupture event. More specifically, the objectives of the Technical Meeting were to review the safety philosophy for the PCP rupture event in pool type LMFR, to assess the structural reliability of the PCP and the probability of rupture under different conditions (with/without in-service inspection), to

Problems specific to the piping of sodium-cooled fast breeder reactors

International Nuclear Information System (INIS)

Vrillon, B.; Befre, J.; Schaller, K.

1975-01-01

A certain number of specific problems arising in connection with the sodium pipes in fast neutron reactors, especially those of large diameter, are presented. The supporting system must be designed to achieve the best compromise among stresses due to weight and various stresses of thermal origin. Large-scale experimental studies carried out on actual elements of the intermediate circuit of the Phenix reactor showed that the circuits can withstand considerable deformation collapse of the walls without danger of leakage. Protection studies against earthquakes are mentionned [fr

TA-2 Water Boiler Reactor Decommissioning Project

International Nuclear Information System (INIS)

Durbin, M.E.; Montoya, G.M.

1991-06-01

This final report addresses the Phase 2 decommissioning of the Water Boiler Reactor, biological shield, other components within the biological shield, and piping pits in the floor of the reactor building. External structures and underground piping associated with the gaseous effluent (stack) line from Technical Area 2 (TA-2) Water Boiler Reactor were removed in 1985--1986 as Phase 1 of reactor decommissioning. The cost of Phase 2 was approximately $623K. The decommissioning operation produced 173 m 3 of low-level solid radioactive waste and 35 m 3 of mixed waste. 15 refs., 25 figs., 3 tabs

Reactor shutdown back-up system

International Nuclear Information System (INIS)

Hirao, Seizo; Sakashita, Motoaki.

1982-01-01

Purpose: To prevent back flow of poison upon injection to a moderator recycling pipeway. Constitution: In a nuclear reactor comprising a moderator recycling system for recycling and cooling moderator through a control rod guide pipe and a rapid poison injection system for rapidly injecting a poison solution at high density into the moderator by way of the same control rod guide pipe as a reactor shutdown back-up system, a mechanism is provided for preventing the back flow of a poison solution at high density into the moderator recycling system upon rapid injection of poison. An orifice provided in the joining pipeway to the control rod guide pipe on the side of the moderator recycling system is utilized as the back flow preventing device for the poison solution and the diameter for the orifice is determined so as to provide a constant ratio between the pressure loss in the control rod guide pipe and the pressure loss in the moderator recycling system pipe line upon usual reactor operation. (Kawakami, Y.)

Small ex-core heat pipe thermionic reactor concept (SEHPTR)

International Nuclear Information System (INIS)

Jacox, M.G.; Bennett, R.G.; Lundberg, L.B.; Miller, B.G.; Drexler, R.L.

1991-01-01

The Idaho National Engineering Laboratory (INEL) has developed an innovative space nuclear power concept with unique features and significant advantages for both Defense and Civilian space missions. The Small Ex-core Heat Pipe Thermionic Reactor (SEHPTR) concept was developed in response to Air Force needs for space nuclear power in the range of 10 to 40 kilowatts. This paper describes the SEHPTR concept and discusses the key technical issues and advantages of such a system

Stress corrosion cracking of nuclear reactor pressure vessel and piping steels

International Nuclear Information System (INIS)

Speidel, M.O.; Magdowski, R.M.

1988-01-01

This paper presents an extensive investigation of stress corrosion cracking of nuclear reactor pressure vessel and piping steels exposed to hot water. Experimental fracture mechanics results are compared with data from the literature and other laboratories. Thus a comprehensive overview of the present knowledge concerning stress corrosion crack growth rates is provided. Several sets of data confirm that 'fast' stress corrosion cracks with growth rates between 10 -8 and 10 -7 m/s and threshold stress intensities around 20 MN m -3/2 can occur under certain conditions. However, it appears possible that specific environmental, mechanical and metallurgical conditions which may prevail in reactors can result in significantly lower stress corrosion crack growth rates. The presently known stress corrosion crack growth rate versus stress intensity curves are discussed with emphasis on their usefulness in establishing safety margins against stress corrosion cracking of components in service. Further substantial research efforts would be helpful to provide a data base which permits well founded predictions as to how stress corrosion cracking in pressure vessels and piping can be reliably excluded or tolerated. It is emphasized, however, that the nucleation of stress corrosion cracks (as opposed to their growth) is difficult and may contribute substantially to the stress corrosion free service behaviour of the overwhelming majority of pressure vessels and pipes. (author)

Stress-assisted, microbial-induced corrosion of stainless steel primary piping and other aging issues at the Omega West Reactor

International Nuclear Information System (INIS)

Andrade, A.

1995-01-01

After the discovery of cooling system leak of about 284 liters per twenty-four (24) hour period, an investigation determined that the 76.2-cm diameter, 33.5-m long stainless-steel (304) OWR delay line was losing water at the same nominal rate. An excavation effort revealed that a circumferential crack, approximately 0.0025 cm in width, extended around the bottom half of the delay line. In addition, other evidence of what appeared to be microcracking and pitting that originated at random nucleated sites around the pipe were also found. Results of destructive analysis and nondestructive testing allowed Los Alamos staff to conclude that the direct cause for the main crack and other pitting resulted from stress-assisted, microbial-induced corrosion of the stainless steel primary piping. The results also indicated that microbial action from bacteria that are normally present in earth can be extremely harmful to stainless- steel piping under certain conditions. Other potential problems that could have also eventually led to a permanent shutdown of the OWR were discussed. These problems, although never encountered nor associated with the current shutdown, were identified in aging studies and are associated with: (1) the water-cooled, bismuth gamma-ray shield and, (2) the aluminum thermal column head seal that prevents reactor vessel water from entering into the graphite-filled thermal column

Piping benchmark problems for the ABB/CE System 80+ Standardized Plant

International Nuclear Information System (INIS)

Bezler, P.; DeGrassi, G.; Braverman, J.; Wang, Y.K.

1994-07-01

To satisfy the need for verification of the computer programs and modeling techniques that will be used to perform the final piping analyses for the ABB/Combustion Engineering System 80+ Standardized Plant, three benchmark problems were developed. The problems are representative piping systems subjected to representative dynamic loads with solutions developed using the methods being proposed for analysis for the System 80+ standard design. It will be required that the combined license licensees demonstrate that their solution to these problems are in agreement with the benchmark problem set. The first System 80+ piping benchmark is a uniform support motion response spectrum solution for one section of the feedwater piping subjected to safe shutdown seismic loads. The second System 80+ piping benchmark is a time history solution for the feedwater piping subjected to the transient loading induced by a water hammer. The third System 80+ piping benchmark is a time history solution of the pressurizer surge line subjected to the accelerations induced by a main steam line pipe break. The System 80+ reactor is an advanced PWR type

Electromagnetic interference produced by power or electrified railway lines on metallic pipe networks

International Nuclear Information System (INIS)

Lucca, G.

1999-01-01

The paper presents an algorithm for the calculation, in the frequency domain, of the induced voltages and currents on a generic metallic pipe network exposed to the electromagnetic interference from a power line or an electrified railway line. By assuming as known the voltages and the currents on the inducing line, the algorithm may be subdivided into the following main steps: a) determination of the ideal electromotive force and current generators to be applied to the induced structure in order to represent the electromagnetic influence from the inducing line; b) modelling of the pipe network by means of a suitable equivalent electric network; c) calculation of voltages and currents on the induced network [it

Pressurizer /Auxiliary Spray Piping Stress Analysis For Determination Of Lead Shielding Maximum Allow Able Load

International Nuclear Information System (INIS)

Setjo, Renaningsih

2000-01-01

Piping stress analysis for PZR/Auxiliary Spray Lines Nuclear Power Plant AV Unit I(PWR Type) has been carried out. The purpose of this analysis is to establish a maximum allowable load that is permitted at the time of need by placing lead shielding on the piping system on class 1 pipe, Pressurizer/Auxiliary Spray Lines (PZR/Aux.) Reactor Coolant Loop 1 and 4 for NPP AV Unit one in the mode 5 and 6 during outage. This analysis is intended to reduce the maximum amount of radiation dose for the operator during ISI ( In service Inspection) period.The result shown that the maximum allowable loads for 4 inches lines for PZR/Auxiliary Spray Lines is 123 lbs/feet

Method of preventing sodium from flowing when pipes of a fast breeder reactor are injured

International Nuclear Information System (INIS)

Nakai, Yasushi; Yamagishi, Yoshiaki; Koga, Tomonari.

1975-01-01

Object: To inject high pressure sodium into an inlet nozzle portion when fluid pressure in the inlet nozzle portion of a core cooling pipe on the inlet side is in an abnormal condition, to thereby quickly and positively prevent the flow of sodium in a high pressure chamber in a reactor vessel, when pipes are injured. Structure: When the core cooling pipe on the inlet side is injured and as a consequence the pressure gage detects an abnormal condition of fluid pressure in the inlet nozzle, the valve is opened to allow high pressure sodium to inject into the inlet nozzle through a high pressure sodium supply pipe, thereby blocking a back-flow of sodium in the high pressure chamber into the core cooling pipe. (Kamimura, M.)

Device for extracting steam or gas from the primary coolant line leading from a reactor pressure vessel to a straight through boiler or from the top primary boiler chamber of a water-cooled nuclear reactor

International Nuclear Information System (INIS)

Schatz, K.

1982-01-01

In such a nuclear reactor, a steam or gas cushion can form when the primary system is refilled, which can cause blocking of the natural circulation or filling of the system in the area of the hot primary coolant pipe or in the top primary boiler chamber. In order to remove such a steam or gas cushion, a ventilation pipe starting from the bend of the primary coolant line is connected to the feed pipe for introducing water into the primary system. The feed pipe is designed on the principle of the vacuum pump in the area of the opening of the ventilation pipe. There is a sub-pressure in the ventilation pipe, which makes it possible to extract the steam or gas. After mixing in the area of the opening, the steam condenses or is distributed with the gas in the primary coolant. (orig.) [de

Model engineering for piping layout of boiling water reactor nuclear station

International Nuclear Information System (INIS)

Tsukada, Koji; Uchiyama, Masayuki; Wada, Takanao; Jibu, Noboru.

1977-01-01

A nuclear power station is made up of a wide variety of equipment, piping, ventilation ducts, conduits, and cable trays, etc. Even if equipment arrangement and piping layout are carefully planned on drawings, troubles such as interference often occur at field installation. Accordingly, it is thought very useful to make thorough examinations with plastic three-dimensional models in addition to drawings in reducing troubles at field, shortening the construction period, and improving economics. Examination with plastic models offers the following features: (1) It permits visual three-dimensional examination. (2) Group thinking and examination is possible. (3) Troubles due to failure to understand complicated drawings can be reduced drastically. Manufacturing a 1/20 scale model of the reactor building of the Tokai No. 2 Power Station of the Japan Atomic Power Co., Hitachi has performed model engineering-solution of interference troubles related to equipment and piping, securing of work space for in-service inspection (ISI), carry-in/installation of various equipment and piping, and determination of the piping route of which only the starting and terminating points were given under the complicated ambient conditions. Success with this procedure has confirmed that model engineering is an effective technique for future plant engineering. (auth.)

Insulated pipe clamp design

International Nuclear Information System (INIS)

Anderson, M.J.; Hyde, L.L.; Wagner, S.E.; Severud, L.K.

1980-01-01

Thin wall large diameter piping for breeder reactor plants can be subjected to significant thermal shocks during reactor scrams and other upset events. On the Fast Flux Test Facility, the addition of thick clamps directly on the piping was undesired because the differential metal temperatures between the pipe wall and the clamp could have significantly reduced the pipe thermal fatigue life cycle capabilities. Accordingly, an insulated pipe clamp design concept was developed. The design considerations and methods along with the development tests are presented. Special considerations to guard against adverse cracking of the insulation material, to maintain the clamp-pipe stiffness desired during a seismic event, to minimize clamp restraint on the pipe during normal pipe heatup, and to resist clamp rotation or spinning on the pipe are emphasized

Diagnosis and on-line displacement monitoring for critical pipe of fossil power plant

Energy Technology Data Exchange (ETDEWEB)

Heo, J. S.; Hyun, J. S. [Korea Electric Power Corporation, Seoul (Korea, Republic of); Heo, J. R.; Lee, S. K.; Cho, S. Y. [Korea South-East Power Co., Ltd., Seoul (Korea, Republic of)

2009-07-01

High temperature steam pipes of fossil power plant are subject to a severe thermal range and usually operates well into the creep range. Cyclic operation of the plant subjects the piping system to mechanical and thermal fatigue mechanisms and poor or malfunctional support assemblies can impose massive loads or stress onto the piping system. In order to prevent the serious damage and failure of the critical pipe system, various inspection methods such as visual inspection, computational analysis and on-line piping displacement monitoring were developed. 3-Dimensional piping displacement monitoring system was developed with using he aluminum alloy rod and rotary encoder type sensors, this system was installed and operated on the 'Y' fossil power plant successfully. It is expected that this study will contribute to the safety of piping system, which could minimize stress and extend the actual life of critical piping.

76 FR 58263 - Kenai Pipe Line Company; Tesoro Alaska Company; Tesoro Logistics Operations, LLC; Notice of...

Science.gov (United States)

2011-09-20

... DEPARTMENT OF ENERGY Federal Energy Regulatory Commission [Docket No. OR11-21-000] Kenai Pipe Line Company; Tesoro Alaska Company; Tesoro Logistics Operations, LLC; Notice of Request for Jurisdictional..., 2011, Kenai Pipe Line Company (KPL), Tesoro Alaska Company (Tesoro Alaska), and Tesoro Logistics, LLC...

Modal analysis of main steam line piping under high energy line break condition

Energy Technology Data Exchange (ETDEWEB)

Kim, Tae-Jin; Kim, Seung Hyun; Je, Sang-Yun; Chang, Yoon-Suk [Kyung Hee University, Yongin (Korea, Republic of)

2016-10-15

If HELB (High Energy Line Break) occurs in NPPs (Nuclear Power Plants), not only environmental effect like release of radioactive material but also secondary structural defects should be considered. Jet impingement phenomenon caused by sudden pipe rupture may lead to severe damage on neighboring safe-related components and other structure. Lots of studies have been conducted to assess dynamic behaviors of the SG and MSL piping while pipe whip restraints and jet impingement shields are taken into account during design stage. Arroyo et al. performed modal analyses of a simple square component to examine the jet impingement phenomenon. Also, structural characteristics were predicted to assure structural integrity against the HELB. In this study, we examined dynamic characteristics of SG and MSL piping in a typical 1000MWe NPP. Simulation was performed by using two commercial computational softwares. In particular, modal analyses were conducted to determine mode shapes and natural frequencies of the structure and maximum displacements. The data obtain from each software were compared and observation was discussed in relation to the jet impingement phenomenon. In this research, modal analyses on the SG and MSL piping were carried out to get natural frequencies, vibration mode shapes and maximum displacements. Thereby, the following key finding was observed. (1) Maximum displacement was calculated at the top of SG outlet nozzle with y-directional bending at the third mode. (2) The differences between two models were respectively 7% in natural frequencies and less than 1% in maximum displacements.

Proceedings of the specialists meeting on experience with thermal fatigue in LWR piping caused by mixing and stratification

International Nuclear Information System (INIS)

1998-01-01

This specialists meeting on experience with thermal fatigue in LWR piping caused by mixing and stratification, was held in June 1998 in Paris. It included five sessions. Session 1: operating experience (7 papers): Historical perspective; EDF experience with local thermohydraulic phenomena in PWRs: impacts and strategies; Thermal fatigue in safety injection lines of French PWRs: technical problems, regulatory requirements, concerns about other areas; US NRC Regulatory perspective on unanticipated thermal fatigue in LWR piping; Failure to the Residual Heat Removal system suction line pipe in Genkai unit 1 caused by thermal stratification cycling; Emergency Core Cooling System pipe crack incident at Tihange unit 1; Two leakages induced by thermal stratification at the Loviisa power plant). Session 2: thermal hydraulic phenomena (5 papers): Thermal stratification in small pipes with respect to fatigue effects and so called 'Banana effect'; Thermal stratification in the surge line of the Korean next generation reactor; Thermal stratification in horizontal pipes investigated in UPTF-TRAM and HDR facilities; Research on thermal stratification in un-isolable piping of reactor pressure boundary; Thermal mixing phenomena in piping systems: 3D numerical simulation and design considerations. Session 3: response of material and structure (5 papers): Fatigue induced by thermal stratification, Results of tests and calculations of the COUFAST model; Laboratory simulation of thermal fatigue cracking as a basis for verifying life models; Thermo-mechanical analysis methods for the conception and the follow up of components submitted to thermal stratification transients; Piping analysis methods of a PWR surge line for stratified flow; The thermal stratification effect on surge lines, The VVER estimation. Session 4: monitoring aspects (4 papers): Determination of the thermal loadings affecting the auxiliary lines of the reactor coolant system in French PWR plants; Expected and

Promethus Hot Leg Piping Concept

International Nuclear Information System (INIS)

AM Girbik; PA Dilorenzo

2006-01-01

The Naval Reactors Prime Contractor Team (NRPCT) recommended the development of a gas cooled reactor directly coupled to a Brayton energy conversion system as the Space Nuclear Power Plant (SNPP) for NASA's Project Prometheus. The section of piping between the reactor outlet and turbine inlet, designated as the hot leg piping, required unique design features to allow the use of a nickel superalloy rather than a refractory metal as the pressure boundary. The NRPCT evaluated a variety of hot leg piping concepts for performance relative to SNPP system parameters, manufacturability, material considerations, and comparison to past high temperature gas reactor (HTGR) practice. Manufacturability challenges and the impact of pressure drop and turbine entrance temperature reduction on cycle efficiency were discriminators between the piping concepts. This paper summarizes the NRPCT hot leg piping evaluation, presents the concept recommended, and summarizes developmental issues for the recommended concept

Experimental basis for parameters contributing to energy dissipation in piping systems

International Nuclear Information System (INIS)

Ibanez, P.; Ware, A.G.

1985-01-01

The paper reviews several pipe testing programs to suggest the phenomena causing energy dissipation in piping systems. Such phenomena include material damping, plasticity, collision in gaps and between pipes, water dynamics, insulation straining, coupling slippage, restraints (snubbers, struts, etc.), and pipe/structure interaction. These observations are supported by a large experimental data base. Data are available from in-situ and laboratory tests (pipe diameters up to about 20 inches, response levels from milli-g's to responses causing yielding, and from excitation wave forms including sinusoid, snapback, random, and seismic). A variety of pipe configurations have been tested, including simple, bare, straight sections and complex lines with bends, snubbers, struts, and insulation. Tests have been performed with and without water and at zero to operating pressure. Both light water reactor and LMFBR piping have been tested

Piping inspection activities at the EPRI NDE Center

International Nuclear Information System (INIS)

Ammirato, F.V.

1988-01-01

Intergranular stress corrosion cracking (IGSCC) in the primary system of boiling water reactors (BWRs) has been a major reliability issue in recent years. BWR pipe cracking was first reported in 1974 with a low percentage of only small-diameter lines affected. However, with increased plant operating time, the number of reported cracking incidents has risen significantly and in 1982 and 1983 included the large-diameter recirculation lines. With the advent of cracking in large-diameter piping, innovative repair remedies were developed, such as weld overlay for repair (WOR). Although these remedies are effective in extending the service life of piping, they also present challenging NDE problems. The EPRI program for improving piping examination has aimed at systematically resolving the difficulties by optimizing techniques and procedures as well as by developing field-qualified automated examination equipment. The EPRI NDE Center's role has been the evaluation and transfer of the technology necessary to address the current piping examination problems of the nuclear utility industry. These activities normally include the following: technology assessment and improvement; validation through demonstrations and field trials; technology transfer reports, workshops, training, and qualification testing; and acquisition of relevant samples. The activities of the NDE Center are discussed

77 FR 21734 - Certain Small Diameter Carbon and Alloy Seamless Standard, Line, and Pressure Pipe From Romania...

Science.gov (United States)

2012-04-11

... DEPARTMENT OF COMMERCE International Trade Administration [A-485-805] Certain Small Diameter Carbon and Alloy Seamless Standard, Line, and Pressure Pipe From Romania: Extension of Time Limit for... diameter carbon and alloy seamless standard, line and pressure pipe from Romania for the period August 1...

Automatic welding processes for reactor coolant pipes used in PWR type nuclear power plant

International Nuclear Information System (INIS)

Hamada, T.; Nakamura, A.; Nagura, Y.; Sakamoto, N.

1979-01-01

The authors developed automatic welding processes (submerged arc welding process and TIG welding process) for application to the welding of reactor coolant pipes which constitute the most important part of the PWR type nuclear power plant. Submerged arc welding process is suitable for flat position welding in which pipes can be rotated, while TIG welding process is suitable for all position welding. This paper gives an outline of the two processes and the results of tests performed using these processes. (author)

Inspection indications, stress corrosion cracks and repair of process piping in nuclear materials production reactors

International Nuclear Information System (INIS)

Louthan, M.R. Jr.; West, S.L.; Nelson, D.Z.

1991-01-01

Ultrasonic inspection of Schedule 40 Type 304 stainless steel piping in the process water system of the Savannah River Site reactors has provided indications of discontinuities in less than 10% of the weld heat affected zones. Pipe sections containing significant indications are replaced with Type 304L components. Post removal metallurgical evaluation showed that the indications resulted from stress corrosion cracking in weld heat-affected zones and that the overall weld quality was excellent. The evaluation also revealed weld fusion zone discontinuities such as incomplete penetration, incomplete fusion, inclusions, underfill at weld roots and hot cracks. Service induced extension of these discontinuities was generally not significant although stress corrosion cracking in one weld fusion zone was noted. One set of UT indications was caused by metallurgical discontinuities at the fusion boundary of an extra weld. This extra weld, not apparent on the outer pipe surface, was slightly overlapping and approximately parallel to the weld being inspected. This extra weld was made during a pipe repair, probably associated with initial construction processes. The two nearly parallel welds made accurate assessment of the UT signal difficult. The implications of these observations to the inspection and repair of process water systems of nuclear reactors is discussed

On estimation of reliability for pipe lines of heat power plants under cyclic loading

International Nuclear Information System (INIS)

Verezemskij, V.G.

1986-01-01

One of the possible methods to obtain a quantitative estimate of the reliability for pipe lines of the welded heat power plants under cyclic loading due to heating-cooling and due to vibration is considered. Reliability estimate is carried out for a common case of loading by simultaneous cycles with different amplitudes and loading asymmetry. It is shown that scattering of the breaking number of cycles for the metal of welds may perceptibly decrease reliability of the welded pipe line

Large butterfly valve design copes with out-of-round pipe

International Nuclear Information System (INIS)

Saar, R.P.

1975-01-01

Two 96 inch circulating water lines at the Trojan reactor were joined to butterfly valves which had to be distorted to conform to the badly out-of-round pipes. Bubble tight seating was achieved by positioning a flexible seat ring after the valve was installed

Nuclear piping criteria for Advanced Light-Water Reactors, Volume 1--Failure mechanisms and corrective actions

International Nuclear Information System (INIS)

Anon.

1993-01-01

This WRC Bulletin concentrates on the major failure mechanisms observed in nuclear power plant piping during the past three decades and on corrective actions taken to minimize or eliminate such failures. These corrective actions are applicable to both replacement piping and the next generation of light-water reactors. This WRC Bulletin was written with the objective of meeting a need for piping criteria in Advanced Light-Water Reactors, but there is application well beyond the LWR industry. This Volume, in particular, is equally applicable to current nuclear power plants, fossil-fueled power plants, and chemical plants including petrochemical. Implementation of the recommendations for mitigation of specific problems should minimize severe failures or cracking and provide substantial economic benefit. This volume uses a case history approach to high-light various failure mechanisms and the corrective actions used to resolve such failures. Particular attention is given to those mechanisms leading to severe piping failures, where severe denotes complete severance, large ''fishmouth'' failures, or long throughwall cracks releasing a minimum of 50 gpm. The major failure mechanisms causing severe failure are erosion-corrosion and vibrational fatigue. Stress corrosion cracking also has been a common problem in nuclear piping systems. In addition thermal fatigue due to mixing-tee and to thermal stratification also is discussed as is microbiologically-induced corrosion. Finally, water hammer, which represents the ultimate in internally-generated dynamic high-energy loads, is discussed

Evaluation of cracking in steam generator feedwater piping in pressurized water reactor plants

International Nuclear Information System (INIS)

Goldberg, A.; Streit, R.D.

1981-05-01

Cracking in feedwater piping was detected near the inlet to steam generators in 15 pressurized water reactor plants. Sections with cracks from nine plants are examined with the objective of identifying the cracking mechanism and assessing various factors that might contribute to this cracking. Using transmission electron microscopy, fatigue striations are observed on replicas of cleaned crack surfaces. Calculations based on the observed striation spacings gave a cyclic stress value of 150 MPa (22 ksi) for one of the major cracks. The direction of crack propagation was invariably related to the piping surface and not to the piping axis. These two factors are consistent with the proposed concept of thermally induced, cyclic, tensile surface stresses and it is concluded that the overriding factor in the cracking problem was the presence of such undocumented cyclic loads

Seismic analysis of nuclear piping system

International Nuclear Information System (INIS)

Shrivastava, S.K.; Pillai, K.R.V.; Nandakumar, S.

1975-01-01

To illustrate seismic analysis of nuclear power plant piping, a simple piping system running between two floors of the reactor building is assumed. Reactor building floor response is derived from time-history method. El Centre earthquake (1940) accelerogram is used for time-history analysis. The piping system is analysed as multimass lumped system. Behaviour of the pipe during the said earthquake is discussed. (author)

Advances in flaw evaluation procedures and acceptance criteria for reactor piping

International Nuclear Information System (INIS)

Gamble, R.M.; Zahoor, A.; Norris, D.M.

1986-01-01

During the past several years, intergranular stress corrosion cracks (IGSCC) have been detected in stainless steel piping in boiling water reactors (BWRs) and have resulted in an increased number of flaw evaluations. To reduce the outage time associated with evaluating IGSCC, various research and ASME code groups have spent significant effort to provide utility personnel with efficient means to detect, classify, and size flaws, and to determine suitability for return to service for flawed stainless steel piping. One of the several nondestructive evaluation technologies that has received considerable attention is fracture mechanics, the discipline that considers the failure of flawed material. Fracture mechanics can be used to answer two key questions concerning return to service of flawed pipe: (a) what is the largest flaw size that can be returned to service and still maintain adequate safety margins at the applied loads, and (b) how much operating time remains before the crack reaches the largest allowable size? The purpose of this paper is to provide an overview of the recently developed ASME code Section XI flaw size evaluation procedure and acceptance criteria for stainless steel piping and their application by BWR owners to efficiently determine if flaws found by nondestructive examination are acceptable for continued service

Advances in flaw evaluation procedures and acceptance criteria for reactor piping

Energy Technology Data Exchange (ETDEWEB)

Gamble, R.M.; Zahoor, A.; Norris, D.M.

1986-01-01

During the past several years, intergranular stress corrosion cracks (IGSCC) have been detected in stainless steel piping in boiling water reactors (BWRs) and have resulted in an increased number of flaw evaluations. To reduce the outage time associated with evaluating IGSCC, various research and ASME code groups have spent significant effort to provide utility personnel with efficient means to detect, classify, and size flaws, and to determine suitability for return to service for flawed stainless steel piping. One of the several nondestructive evaluation technologies that has received considerable attention is fracture mechanics, the discipline that considers the failure of flawed material. Fracture mechanics can be used to answer two key questions concerning return to service of flawed pipe: (a) what is the largest flaw size that can be returned to service and still maintain adequate safety margins at the applied loads, and (b) how much operating time remains before the crack reaches the largest allowable size. The purpose of this paper is to provide an overview of the recently developed ASME code Section XI flaw size evaluation procedure and acceptance criteria for stainless steel piping and their application by BWR owners to efficiently determine if flaws found by nondestructive examination are acceptable for continued service.

Calculation of Local Stress and Fatigue Resistance due to Thermal Stratification on Pressurized Surge Line Pipe

Science.gov (United States)

Bandriyana, B.; Utaja

2010-06-01

Thermal stratification introduces thermal shock effect which results in local stress and fatique problems that must be considered in the design of nuclear power plant components. Local stress and fatique calculation were performed on the Pressurize Surge Line piping system of the Pressurize Water Reactor of the Nuclear Power Plant. Analysis was done on the operating temperature between 177 to 343° C and the operating pressure of 16 MPa (160 Bar). The stagnant and transient condition with two kinds of stratification model has been evaluated by the two dimensional finite elements method using the ANSYS program. Evaluation of fatigue resistance is developed based on the maximum local stress using the ASME standard Code formula. Maximum stress of 427 MPa occurred at the upper side of the top half of hot fluid pipe stratification model in the transient case condition. The evaluation of the fatigue resistance is performed on 500 operating cycles in the life time of 40 years and giving the usage value of 0,64 which met to the design requirement for class 1 of nuclear component. The out surge transient were the most significant case in the localized effects due to thermal stratification.

MATHEMATICAL MODEL OF POWER CONSUMPTION FOR SOME OIL PIPE-LINE SECTIONS WITH POOR OPERATIONAL STABILITY

Directory of Open Access Journals (Sweden)

J. N. Kolesnik

2005-01-01

Full Text Available Mathematical model of power consumption for technologically completed and non-completed oil pipe-line sections with poor operational stability has been developed on the basis of daily indices concerning oil transportation regimes. The model permits to take into account tendencies in power consumption under various time prediction cycles and ranges of oil freight turnover, changes in the bulk and characteristics of the transported oil, configuration and design parameters of oil pipe-line.

Screening dynamic evaluation of SRS cooling water line

International Nuclear Information System (INIS)

Bezler, P.; Shteyngart, S.; Breidenbach, G.

1991-01-01

The production reactors at the Savannah River Site (SRS) have been shut down due to perceived safety concerns. A major concern is the seismic integrity of the plant. A comprehensive program is underway to assess the seismic capacity of the existing systems and components and to upgrade them to acceptable levels. The evaluation of the piping systems at the SRS is a major element of this program. Many of the piping systems at the production reactors were designed without performing dynamic analyses. Instead their design complied with good design practice for dead weight supported systems with proper accommodation of thermal expansion effects. In order to gain some insight as to the seismic capacity of piping installed in this fashion, dynamic analyses were performed for some lines. Since the piping was not seismically supported, the evaluations involved various approximations and the results are only used as a screening test of seismic adequacy. In this paper, the screening evaluations performed for the raw water inlet line are described. This line was selected for evaluation since it was considered typical of the smaller diameter piping systems at the plant. It is a dead weight supported system made up of a run of small diameter piping which extends for great distances over many dead weight supports and through wall penetrations. The results of several evaluations for the system using different approximations to represent the support system are described. 2 figs., 4 tabs

Causes of pipe ruptures in distribution lines. Evaluation of long-term observations in a metropolitan pipe network

Energy Technology Data Exchange (ETDEWEB)

Kottmann, A

1978-01-01

Pipe ruptures and their causes are examined from the viewpoints of pipe material, corrosion, traffic, internal pressure, air temperature, ground temperature, ground frost, gas or water temperature, and ground moisture level. The examination relies on 17 years of statistics (1958-74) from (1) Technische Werke der Stadt Stuttgart AG on 11,986 pipe ruptures and (2) German weather-service data on ground-moisture readings at depths down to 80 in. in the Stuttgart area. Faced with replacing up to 280 miles (450 km) of cast-iron gas-distribution lines that seemed extraordinarily prone to rupture (company records showed at least 20 breaks/month) after the conversion to natural gas, TWS authorized this study to determine the boundary conditions that make cast-iron pipe susceptible to fracture, thus minimizing the extent of the replacement program. The investigation showed that corrosion had only a slight effect upon cracking. No significant effect was found for any of the following: temperature-caused changes in material properties, internal pressure or pressure changes, fluctuations in gas temperature, changes in air temperature, and summertime changes in ground temperature. Stress loading by heavy traffic, however, doubled the fracture incidence.

Development of forging technology for PWR primary piping

International Nuclear Information System (INIS)

Morin, F.; Badeau, J.P.; Lambs, R.

1996-01-01

The purpose of this presentation is to give information on the changes in the design and manufacture of Primary Piping for electronuclear boilers of the Pressurized Water Reactor type (PWR) which has resulted in the making of one-piece forged lines including stub pipes and arcs. The optimization of these items is aimed at improving the life of the new power stations as well as guaranteeing their safety, while reducing inspection and maintenance requirements in service. The demonstration of the manufacturing feasibility has just been completed. It has taken material form in the installation, on the CIVAUX 1 section, of the first one-piece cold leg in the world. It will shortly be followed by the installation on the CIVAUX 2 section of a complete loop of bent forged pipes. Therefore, this new know-how is going to be incorporated in the French Rules (RCC-M) and can be directly taken into consideration both in the next work to be done and in the design and definition of a future nuclear reactor

Nuclear reactor

International Nuclear Information System (INIS)

Schabert, H.P.; Weber, R.; Bauer, A.

1975-01-01

The refuelling of a PWR power reactor of about 1,200 MWe is performed by a transport pipe in the containment leading from an external to an internal fuel pit. A wagon to transport the fuel elements can go from a vertical loading position to an also vertical deloading position in the inner fuel pit via guide rollers. The necessary horizontal movement is effected by means of a cable line through the transport pipe which is inclined at least 10 0 . Gravity thus helps in the movement to the deloading position. The cable line with winch is fastened outside the containment. Swivelling devices tip the wagon from the horizontal to the vertical position or vice versa. Loading and deloading are done laterally. (TK/LH) [de

Model for cobalt 60/58 deposition on primary coolant piping in a boiling water reactor

International Nuclear Information System (INIS)

Dehollander, W.R.

1979-01-01

A first principles model for deposition of radioactive metals into the corrosion films of primary coolant piping is proposed. It is shown that the predominant mechanism is the inclusion of the radioactive species such as Cobalt 60 into the spinel structure of the corrosion film during the act of active corrosion. This deposition can occupy only a defined fraction of the available plus 2 valence sites of the spinel. For cobalt ions, this ratio is roughly 4.6 x 10 -3 of the total iron sites. Since no distinction is made between Cobalt 60, Cobalt 58, and Cobalt 59 in this process, the radioactivity associated with this inclusion is a function of the ratio of the radioactive species to the nonradioactive species in the water causing the corrosion of the pipe metal. The other controlling parameter is the corrosion rate of the pipe material. This can be a function of time, for example, and it shown that freshly descaled metal when exposed to the cobalt containing water can incorporate as much as 10 x 10 -3 cobalt ions per iron atom in the initial corrosion period. This has implications for the problem of decontaminating nuclear reactor piping. Equations and selected observations are presented without reference to any specifically identified reactor or utility, so as to protect any proprietary interest

Behavior study on Na heat pipe in passive heat removal system of new concept molten salt reactor

International Nuclear Information System (INIS)

Wang Chenglong; Tian Wenxi; Su Guanghui; Zhang Dalin; Wu Yingwei; Qiu Suizheng

2013-01-01

The high temperature Na heat pipe is an effective device for transporting heat, which is characterized by remarkable advantages in conductivity, isothermally and passively working. The application of Na heat pipe on passive heat removal system of new concept molten salt reactor (MSR) is significant. The transient performance of high temperature Na heat pipe was simulated by numerical method under the MSR accident. The model of the Na heat pipe was composed of three conjugate heat transfer zones, i.e. the vapor, wick and wall. Based on finite element method, the governing equations were solved by making use of FORTRAN to acquire the profiles of the temperature, velocity and pressure for the heat pipe transient operation. The results show that the high temperature Na heat pipe has a good performance on operating characteristics and high heat transfer efficiency from the frozen state. (authors)

Failure analysis of cracked head spray piping from the Dresden Unit 2 Boiling Water Reactor

International Nuclear Information System (INIS)

Diercks, D.R.; Dragel, G.M.

1983-07-01

Several sections of Type 304 stainless steel head spray piping, 6.25 cm (2.5 in.) in diameter, from the Dresden Unit 2 Boiling Water Reactor were examined to determine the nature and causes of coolant leakages detected during hydrostatic tests. Extensive pitting was observed on the outside surface of the piping, and three cracks, all located at a helical stripe apparently rubbed onto the outer surface of the piping, were also noted. Metallographic examination revealed that the cracking had initiated at the outer surface of the pipe, and showed it to be transgranular and highly branched, characteristic of chloride stress corrosion cracking. The surface pitting also appeared to have been caused by chlorides. A scanning electron microprobe x-ray analysis of the corrosion product in the cracks confirmed the presence of chlorides and also indicated the presence of calcium

78 FR 78350 - Houston Pipe Line Company, LP; Notice of Intent to Prepare an Environmental Assessment for the...

Science.gov (United States)

2013-12-26

... DEPARTMENT OF ENERGY Federal Energy Regulatory Commission [Docket No. CP14-13-000] Houston Pipe Line Company, LP; Notice of Intent to Prepare an Environmental Assessment for the Proposed 24-Inch... Hidalgo County, Texas by Houston Pipe Line Company, LP (HPL). The Commission will use this EA in its...

Probabilistic evaluation of main coolant pipe break indirectly induced by earthquakes Savannah River Project L and P Reactors

International Nuclear Information System (INIS)

Short, S.A.; Wesley, D.A.; Awadalla, N.G.; Kennedy, R.P.

1989-01-01

A probabilistic evaluation of seismically-induced indirect pipe break for the Savannah River Project (SRP) L- and P-Reactor main coolant (process water) piping has been conducted. Seismically-induced indirect pipe break can result primarily from: (1) failure of the anchorage of one or more of the components to which the pipe is anchored; or (2) failure of the pipe due to collapse of the structure. the potential for both types of seismically-induced indirect failures was identified during a seismic walkdown of the main coolant piping. This work involved: (1) identifying components or structures whose failure could result in pipe failure; (2) developing seismic capacities or fragilities of these components; (3) combining component fragilities to develop plant damage state fragilities; and (4) convolving the plant seismic fragilities with a probabilistic seismic hazard estimate for the site in order to obtain estimates of seismic risk in terms of annual probability of seismic-induced indirect pipe break

Structural analyses on piping systems of sodium reactors. 2. Eigenvalue analyses of hot-leg pipelines of large scale sodium reactors

International Nuclear Information System (INIS)

Furuhashi, Ichiro; Kasahara, Naoto

2002-01-01

Two types of finite element models analyzed eigenvalues of hot-leg pipelines of a large-scale sodium reactor. One is a beam element model, which is usual for pipe analyses. The other is a shell element model to evaluate particular modes in thin pipes with large diameters. Summary of analysis results: (1) A beam element model and a order natural frequency. A beam element model is available to get the first order vibration mode. (2) The maximum difference ratio of beam mode natural frequencies was 14% between a beam element model with no shear deformations and a shell element model. However, its difference becomes very small, when shear deformations are considered in beam element. (3) In the first order horizontal mode, the Y-piece acts like a pendulum, and the elbow acts like the hinge. The natural frequency is strongly affected by the bending and shear rigidities of the outer supporting pipe. (4) In the first order vertical mode, the vertical sections of the outer and inner pipes moves in the axial-directional piston mode, the horizontal section of inner pipe behaves like the cantilever, and the elbow acts like the hinge. The natural frequency is strongly affected by the axial rigidity of outer supporting pipe. (5) Both effective masses and participation factors were small for particular shell modes. (author)

The role of heat pipes in intensified unit operations

International Nuclear Information System (INIS)

Reay, David; Harvey, Adam

2013-01-01

Heat pipes are heat transfer devices that rely, most commonly, on the evaporation and condensation of a working fluid contained within them, with passive pumping of the condensate back to the evaporator. They are sometimes referred to as ‘thermal superconductors’ because of their exceptionally high effective thermal conductivity (substantially higher than any metal). This, together with several other characteristics make them attractive to a range of intensified unit operations, particularly reactors. The majority of modern computers deploy heat pipes for cooling of the CPU. The application areas of heat pipes come within a number of broad groups, each of which describes a property of the heat pipe. The ones particularly relevant to chemical reactors are: i. Separation of heat source and sink. ii. Temperature flattening, or isothermalisation. iii. Temperature control. Chemical reactors, as a heat pipe application area, highlight the benefits of the heat pipe based on isothermalisation/temperature flattening device and on being a highly effective heat transfer unit. Temperature control, done passively, is also of relevance. Heat pipe technology offers a number of potential benefits to reactor performance and operation. The aim of increased yield of high purity, high added value chemicals means less waste and higher profitability. Other intensified unit operations, such as those employing sorption processes, can also profit from heat pipe technology. This paper describes several variants of heat pipe and the opportunities for their use in intensified plant, and will give some current examples. -- Highlights: ► Heat pipes – thermal superconductors – can lead to improved chemical reactor performance. ► Isothermalisation within a reactor vessel is an ideal application. ► The variable conductance heat pipe can control reaction temperatures within close limits. ► Heat pipes can be beneficial in intensified reactors

Elevated temperature mechanical properties of line pipe steels

Science.gov (United States)

Jacobs, Taylor Roth

The effects of test temperature on the tensile properties of four line pipe steels were evaluated. The four materials include a ferrite-pearlite line pipe steel with a yield strength specification of 359 MPa (52 ksi) and three 485 MPa (70 ksi) yield strength acicular ferrite line pipe steels. Deformation behavior, ductility, strength, strain hardening rate, strain rate sensitivity, and fracture behavior were characterized at room temperature and in the temperature range of 200--350 °C, the potential operating range for steels used in oil production by the steam assisted gravity drainage process. Elevated temperature tensile testing was conducted on commercially produced as-received plates at engineering strain rates of 1.67 x 10 -4, 8.33 x 10-4, and 1.67 x 10-3 s-1. The acicular ferrite (X70) line pipe steels were also tested at elevated temperatures after aging at 200, 275, and 350 °C for 100 h under a tensile load of 419 MPa. The presence of serrated yielding depended on temperature and strain rate, and the upper bound of the temperature range where serrated yielding was observed was independent of microstructure between the ferrite-pearlite (X52) steel and the X70 steels. Serrated yielding was observed at intermediate temperatures and continuous plastic deformation was observed at room temperature and high temperatures. All steels exhibited a minimum in ductility as a function of temperature at testing conditions where serrated yielding was observed. At the higher temperatures (>275 °C) the X52 steel exhibited an increase in ductility with an increase in temperature and the X70 steels exhibited a maximum in ductility as a function of temperature. All steels exhibited a maximum in flow strength and average strain hardening rate as a function of temperature. The X52 steel exhibited maxima in flow strength and average strain hardening rate at lower temperatures than observed for the X70 steels. For all steels, the temperature where the maximum in both flow

Impact of inservice inspection on the reliability of nuclear piping

International Nuclear Information System (INIS)

Woo, H.H.

1983-12-01

The reliability of nuclear piping is a function of piping quality as fabricated, service loadings and environments, plus programs of continuing inspection during operation. This report presents the results of a study of the impact of inservice inspection (ISI) programs on the reliability of specific nuclear piping systems that have actually failed in service. Two major factors are considered in the ISI programs: one is the capability of detecting flaws; the other is the frequency of performing ISI. A probabilistic fracture mechanics model issued to estimate the reliability of two nuclear piping lines over the plant life as functions of the ISI programs. Examples chosen for the study are the PWR feedwater steam generator nozzle cracking incident and the BWR recirculation reactor vessel nozzle safe-end cracking incident

Nonlinear dynamic response analysis in piping system for a loss of coolant accident in primary loop of pressurized water reactor

International Nuclear Information System (INIS)

Zhang Xiwen; He Feng; Hao Pengfei; Wang Xuefang

2000-01-01

Based on the elaborate force and moment analysis with characteristics method and control-volume integrating method for the piping system of primary loop under pressurized water reactor' loss of coolant accident (LOCA) conditions, the nonlinear dynamic response of this system is calculated by the updated Lagrangian formulation (ADINA code). The piping system and virtual underpinning are specially processed, the move displacement of the broken pipe with time is accurately acquired, which is very important and useful for the design of piping system and virtual underpinning

Study on feasibility of replacing 321 with 316LN stainless steel for main reactor coolant pipe material

International Nuclear Information System (INIS)

Luo Yijun

2013-01-01

The metallurgical, physical and mechanical performance, and the corrosion and welding properties of 00Cr17Ni12Mo2 (controlled Nitrogen, ANSI316LN) and 0Cr18Ni10Ti (ANSI321SS) for main pipe material were analyzed comparatively in this paper. The feasibility of 316LN pipe material manufacturing was studied too. The analysis results showed that under the operation condition of the nuclear reactor, the general properties of 316LN are better than that of 321SS. Therefore, 316LN could be used for main pipe material, replacing 321SS. (authors)

Subprogram Calculating The Distance Between Pipe And Plane For Automatic Piping System Design

International Nuclear Information System (INIS)

Satmoko, Ari

2001-01-01

DISTLNPL subprogram was created using Auto LISP software. This subprogram is planned to complete CAPD (Computer Aided Piping Design) software being developed. The CAPD works under the following method: suggesting piping system line and evaluating whether any obstacle allows the proposed line to be constructed. DISTLNPL is able to compute the distance between pipe and any equipment having plane dimension such as wall, platform, floors, and so on. The pipe is modeled by using a line representing its axis, and the equipment is modeled using a plane limited by some lines. The obtained distance between line and plane gives information whether the pipe crosses the equipment. In the case of crashing, the subprogram will suggest an alternative point to be passed by piping system. So far, DISTLNPL has not been able to be accessed by CAPD yet. However, this subprogram promises good prospect in modeling wall, platform, and floors

Damage mechanism of piping welded joints made from austenitic Steel for the type RBMK reactor

International Nuclear Information System (INIS)

Karzov, G.; Timofeev, B.; Gorbakony, A.; Petrov, V.; Chernaenko, T.

1999-01-01

In the process of operation of RBMK reactors the damages were taking place on welded piping, produced from austenitic stainless steel of the type 08X18H10T. The inspection of damaged sections in piping has shown that in most cases crack-like defects are of corrosion and mechanical character. The paper considers in details the reasons of damages appearance and their development for this type of welded joints of downcomers 325xl6 mm, which were fabricated from austenitic stainless steel using TlG and MAW welding methods. (author)

Demonstration of leak-before-break in Japan Sodium cooled Fast Reactor (JSFR) pipes

International Nuclear Information System (INIS)

Wakai, Takashi; Machida, Hideo; Yoshida, Shinji; Xu, Yang; Tsukimori, Kazuyuki

2014-01-01

This paper describes the leak-before-break (LBB) assessment procedure applicable to Japan Sodium cooled Fast Reactor (JSFR) pipes made of modified 9Cr–1Mo steel. For the sodium pipes of JSFR, the continuous leak monitoring will be adopted as an alternative to a volumetric test of the weld joints under conditions that satisfy LBB. Firstly, a LBB assessment flowchart eliminating uncertainty resulted from small scale leakage, such as self plugging phenomenon and influence of crack surface roughness on leak rate, was proposed. Secondly, a rational unstable fracture assessment technique, taking the compliance changing with crack extension into account, was also proposed. Thirdly, a crack opening displacement (COD) assessment technique was developed, because COD assessment method applicable to JSFR pipes – thin wall and small work hardening material – had not been proposed yet. In addition, fracture toughness tests were performed using compact tension (CT) specimens to obtain the fracture toughness, J IC , and the crack growth resistance (J–R) curve at elevated temperature. Finally, by using the flowchart, proposed techniques and collected data, LBB assessment for the primary sodium pipes of JSFR was conducted. As a result, LBB aspect was successfully demonstrated with sufficient margins

Bypass line assisted start-up of a loop heat pipe with a flat evaporator

International Nuclear Information System (INIS)

Boo, Joon Hong; Jung, Eui Guk

2009-01-01

Loop heat pipes often experience start-up problems especially under low thermal loads. A bypass line was installed between the evaporator and the liquid reservoir to alleviate the difficulties associated with start-up of a loop heat pipe with flat evaporator. The evaporator and condenser had dimensions of 40 mm (W) by 50 mm (L). The wall and tube materials were stainless steel and the working fluid was methanol. Axial grooves were provided in the flat evaporator to serve as vapor passages. The inner diameters of liquid and vapor transport lines were 2 mm and 4 mm, respectively, and the length of the two lines was 0.5 m each. The thermal load range was up to 130 W for horizontal alignment with the condenser temperature of 10 .deg. C. The experimental results showed that the minimum thermal load for start-up was lowered by 37% when the bypass line was employed

Solar chemical heat pipe

International Nuclear Information System (INIS)

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

1991-08-01

The performance of a solar chemical heat pipe was studied using CO 2 reforming of methane as a vehicle for storage and transport of solar energy. The endothermic reforming reaction was carried out in an Inconel reactor, packed with a Rh catalyst. The reactor was suspended in an insulated box receiver which was placed in the focal plane of the Schaeffer Solar Furnace of the Weizman Institute of Science. The exothermic methanation reaction was run in a 6-stage adiabatic reactor filled with the same Rh catalyst. Conversions of over 80% were achieved for both reactions. In the closed loop mode the products from the reformer and from the metanator were compressed into separate storage tanks. The two reactions were run either separately or 'on-line'. The complete process was repeated for over 60 cycles. The overall performance of the closed loop was quite satisfactory and scale-up work is in progress in the Solar Tower. (authors). 35 refs., 2 figs

Evaluation of Bamboo Porous Pipe as Line Source Emitter in Trickle ...

African Journals Online (AJOL)

This paper attempts to evaluate the use of bamboo as porous pipe (line source) emitter in trickle irrigation at the Cross River University of Technology Teaching and Research Farm Obubra. Two sets of bamboo laterals: opened and plugged ends were used for the trial. The experiment was conducted using four different ...

evaluation of bamboo porous pipe as line source emitter in trickle ...

African Journals Online (AJOL)

CHRISTY

This paper attempts to evaluate the use of bamboo as porous pipe (line source) emitter in trickle irrigation at the Cross River University of Technology Teaching and Research Farm Obubra. Two sets of bamboo laterals: opened and plugged ends were used for the trial. The experiment was conducted using four different ...

Structural evaluation report of piping and support structure for design-changed hot-water layer system

International Nuclear Information System (INIS)

Ryu, Jeong Soo

1998-05-01

After hot-water layer system had been installed, the verification tests to reduce the radiation level at the top of reactor pool were performed many times. The major goal of this report is to assess the structural integrity on the piping and the support structures of design-changed hot-water layer system. The piping stress analysis was performed by using ADLPIPE program for the pump suction line and the pump discharge line subjected to dead weight, pressure, thermal expansion and seismic loadings. The stress analysis of the support structure was carried out using the reaction forces obtained from the piping stress analysis. The results of structural evaluation for the pipings and the support structures showed that the structural acceptance criteria were satisfied, in compliance with ASME, subsection ND for the piping and subsection NF for the support structures. Therefore based on the results of the analysis and the design, the structural integrity on the piping and the support structures of design-changed hot-water system was proved. (author). 9 refs., 9 tabs., 14 figs

Characteristics of DC electrical braking method of the gas circulator to limit the temperature rise at the heat transfer pipes in the HTTR

International Nuclear Information System (INIS)

Kawasaki, K.; Saito, K.; Iyoku, T.

2001-01-01

In the safety evaluation of a High Temperature Engineering Test Reactor (HTTR), it must be confirmed that the core has no chance to be damaged and the barrier against the FP release is designed properly not to be affecting the influence of radiation around the reactor site. Especially the maximum temperature of the reactor pressure boundary such as the heat transfer pipes of pressurized water cooler (PWC) must not exceed the permissible values under an anticipated accident such as pipe of rupture in PWC. A requirement for the gas circulator which circulates helium gas in the primary cooling line and the secondary cooling line, is to be braked within 10 seconds by an electrical braking method after the HTTR reactor has scrammed under the accident in PWC. The reason is that the temperature rise of the heat transfer pipe at PWC has to be suppressed when the gas circulator has stopped, the revolution of the gas circulator decreases like the free coast down so that it takes about 90 seconds to be zero and the temperature rise of the pipe in the PWC exceeds the permissible value. By braking within 10 secs., the temperature of the pipe in the PWC reaches about 368 deg. C, less than the permissible value. Using a simplified equivalent circuit of an induction motor, braking time analysis was performed with obtained electrical resistance and inductance. The obtained braking time is about 10 secs., showing close agreement with analysis values. (author)

Selective Method for the Determination of Manganese in End-fitting of Spoolable Reinforced Plastic Line Pipe for Petroleum Industries

Science.gov (United States)

Shao, Xiaodong; Zhang, Dongna; Li, Houbu; Cai, Xuehua

2017-10-01

The fact that spoolable reinforced plastic line pipe is more flexible and spoolable than steel, and is also much lighter, means that it can becarried and deployedfrom smaller vessels and managed more easily. It was well known that manganese is an important element in end-fitting of spoolable reinforced plastic line pipe. In this paper, a simple spectrophotometric method was described for the determination of manganese in end-fitting of spoolable reinforced plastic line pipe. The method was based on the oxidation-reduction reaction between ammonium persulfate and manganese(II) producing manganese(VII) in the presence of silver nitrate as a catalyst. The characteristic wavelength of maximum absorption of manganese(VII) was obtained locating at 530 nm. Under the optimum reaction conditions the absorption value was proportional to the concentration of manganese in the range of 0.50%˜1.80% (R2 = 0.9997), and the relative standard deviation was less than 3.0% (n=5). The proposed method was applied successfully to determine manganese in end-fitting of spoolable reinforced plastic line pipe samples.

Failure behaviour of a piping system with a circumferentially orientated flaw

International Nuclear Information System (INIS)

Mikkola, T.P.J.; Diem, H.; Blind, D.; Hunger, H.

1987-01-01

The experiments were conducted on the recently installed feedwater line of the HDR reactor in Kahl. The investigations were focused on analysing both the crack propagation of a circumferentially flowed pipe under the influence of corrosion and cyclic load, together with the pipeline's subsequent failure behaviour. The experimental conditions were selected in a manner representing those which can, for example, prevait during start-up or shut-down of reactor. To this aim, the pipes were internally stressed with high pressure and temperature oxygenic water in conjunction with an externally applied bending moment. The investigations are supplemented by elastic-plastic triaxial finite element (FE) calculations for various assumed crack configurations, both prior to and following the experiments, thus granting a fracture-mechanical assessment of the structural behaviour. (orig./DG) [de

On detection and automatic tracking of butt weld line in thin wall pipe welding by a mobile robot with visual sensor

International Nuclear Information System (INIS)

Suga, Yasuo; Ishii, Hideaki; Muto, Akifumi

1992-01-01

An automatic pipe welding mobile robot system with visual sensor was constructed. The robot can move along a pipe, and detect the weld line to be welded by visual sensor. Moreover, in order to make an automatic welding, the welding torch can track the butt weld line of the pipes at a constant speed by rotating the robot head. Main results obtained are summarized as follows: 1) Using a proper lighting fixed in front of the CCD camera, the butt weld line of thin wall pipes can be recongnized stably. In this case, the root gap should be approximately 0.5 mm. 2) In order to detect the weld line stably during moving along the pipe, a brightness distribution measured by the CCD camera should be subjected to smoothing and differentiating and then the weld line is judged by the maximum and minimum values of the differentials. 3) By means of the basic robot system with a visual sensor controlled by a personal computer, the detection and in-process automatic tracking of a weld line are possible. The average tracking error was approximately 0.2 mm and maximum error 0.5 mm and the welding speed was held at a constant value with error of about 0.1 cm/min. (author)

Piping reliability model development, validation and its applications to light water reactor piping

International Nuclear Information System (INIS)

Woo, H.H.

1983-01-01

A brief description is provided of a three-year effort undertaken by the Lawrence Livermore National Laboratory for the piping reliability project. The ultimate goal of this project is to provide guidance for nuclear piping design so that high-reliability piping systems can be built. Based on the results studied so far, it is concluded that the reliability approach can undoubtedly help in understanding not only how to assess and improve the safety of the piping systems but also how to design more reliable piping systems

The effect of 25 years of oil field flow line service on epoxy fiberglass pipe

International Nuclear Information System (INIS)

Oswald, K.J.

1988-01-01

Glass fiber reinforced epoxy and vinyl ester piping systems have been used for over 35 years to control corrosion problems in oil fields and chemical and industrial plants and many case histories have been reported to document the successful performances of fiberglass reinforced thermosetting plastics in a wide range of corrosive services. This information is reinforced by laboratory test data from flat laminates and pipe exposed to numerous chemicals and mixtures of chemicals, but little has been published to document the effect of long-term, in-service exposure on fiberglass equipment. The purpose of this paper is to help to fill this void by comparing data from physical testing of pipe removed from successful corrosive service applications with data obtained from the same type of pipe at the time of manufacture. The information supplied in these papers represents only a few of the successful applications of filament wound epoxy and vinyl ester pipe as it is difficult to obtain permission to remove pipe from an operating line

US NRC research on the integrity of piping in nuclear reactor primary systems

International Nuclear Information System (INIS)

Serpan, C.Z. Jr.

1983-01-01

This paper has attempted to provide a ''snapshot'' of the activities underway in NRC on the subject of LWR piping integrity as of the summer and fall of 1983. The paper is necessarily vague on certain topics of policy because they are either under review or are under development and the outcome cannot be accurately forecast at this time. Particularly in the area of BWR pipe cracking, events are very rapid so that positions and actions described in this paper may well be obsolete by the time it is published. Nevertheless, the activities and positions are as accurate as possible at the time of writing. Certainly the longer-range aspects of the research program represent the current direction and intent of NRC; nevertheless, as results come in and actions occur in the licensing and regulation arena of operating reactors, the emphasis of the research programs will necessarily shift to accommodate them so as to remain as relevant as possible. Thus, this paper is useful to show the intentions of NRC in the area of research for LWR piping, and it is also useful to document the status of the regulations on piping for which the research is being performed. (orig.)

Nuclear reactor, reactor core thereof, and device for constituting the reactor

International Nuclear Information System (INIS)

Takiyama, Masashi.

1994-01-01

A reactor core is constituted by charging coolants (light water) in a reactor pressure vessel and distributing fuel assemblies, reflecting material sealing pipes, moderator (heavy water and helium gas) sealing pipes, and gas sealing pipes therein. A fuel guide tube is surrounded by a cap and the gap therebetween is made hollow and filled with coolant steams. The cap is supported by a baffle plate. The moderator sealing pipe is disposed in a flow channel of coolants in adjacent with the cap. The position of the moderator sealing tube in the reactor core is controlled by water stream from a hydraulic pump with a guide tube extending below the baffle plate being as a guide. Then, the position of the moderator sealing tube is varied to conduct power control, burnup degree compensation, and reactor shut down. With such procedures, moderator cooling facility is no more necessary to simplify the structure. Further, heat generated from the moderator is transferred to the coolants thereby improving heat efficiency of the reactor. (I.N.)

Heat pipe

International Nuclear Information System (INIS)

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

1986-01-01

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

PIPE STRESS and VERPIP codes for stress analysis and verifications of PEC reactor piping

International Nuclear Information System (INIS)

Cesari, F.; Ferranti, P.; Gasparrini, M.; Labanti, L.

1975-01-01

To design LMFBR piping systems following ASME Sct. III requirements unusual flexibility computer codes are to be adopted to consider piping and its guard-tube. For this purpose PIPE STRESS code previously prepared by Southern-Service, has been modified. Some subroutine for detailed stress analysis and principal stress calculations on all the sections of piping have been written and fitted in the code. Plotter can also be used. VERPIP code for automatic verifications of piping as class 1 Sct. III prescriptions has been also prepared. The results of PIPE STRESS and VERPIP codes application to PEC piping are in section III of this report

A study of the long-range inspection method for on-line monitoring of pipes in nuclear power plants

International Nuclear Information System (INIS)

Eom, Heung Seop; Lim, Sa Hoe; Kim, Jae Hee; Kim, Young H.; Song, Sung Jin

2005-01-01

Deployment of an advanced on-line monitoring of the component integrity offers the prospect of an improved performance, enhanced safety, and reduced overall cost for nuclear power plants (NPPs). Also ultrasonic guided ultrasonic wave has been known as one of the promising techniques that could be utilized for on-line monitoring, because it enables us to undertake a long-range inspection of structures such as plates and pipes. The present work is aimed at developing a new method using ultrasonic guided waves for the on-line monitoring of pipes. For this purpose we fabricated the necessary hardware and carried out transmitter tuning, group velocity measurement, receiver tuning, and mode identification. Finally we carried out an experiment on a long-range inspection with the developed hardware and the techniques. In the experiment, we could detect the flaws at a distance of about 20M from the transmitter, and we could verify the possibility of using the developed hardware and techniques for on-line monitoring of pipes in NPPs

Remote mechanized equipment for the repair and replacement of boiling water reactor recirculation loop piping

International Nuclear Information System (INIS)

Mauser, D.; Busch, D.F.

1983-01-01

Equipment has been assembled for the remote repair or replacement of boiling water reactor nuclear plant piping in the diameter range of 4 to 28 inches (10-71 cm). The objectives of this program were to produce high-quality pipe welds, reduce plant downtime, and reduce man-rem exposure. The repair strategy was to permit repair personnel to install and check out the repair subsystems and then leave the radiation zone allowing the operations to be conducted at a distance of up to 300 feet (91 m) from the operator. The complete repair system comprises subsystems for pipe severing, dimensional gaging, joint preparation, counterboring, welding, postweld nondestructive inspection (conceptual design), and audio, electronic, and visual monitoring of all operations. Components for all subsystems, excluding those for postweld nondestructive inspection, were purchased and modified as needed for integration into the repair system. Subsystems were designed for two sizes of Type 304 stainless steelpipe. For smaller, 12-inch-diameter (30.5 cm) pipe, severing is accomplished by a power hack saw and joint preparation and counterboring by an internally mounted lathe. The 22-inch-diameter (56 cm) pipe is severed, prepared, and counterbored using an externally mounted, single-point machining device. Dimensional gaging is performed to characterize the pipe geometry relative to a fixed external reference surface, allowing the placement of the joint preparation and the counterbore to be optimized. For both pipe sizes, a track-mounted gas tungsten-arc welding head with filler wire feed is used

Prevention of biofouling and biocorrosion in reactor systems

International Nuclear Information System (INIS)

Mathur, A.K.; Shivananda, S.R.

1995-01-01

Formaldehyde even at 500 μl/l concentration can prevent the growth of bacteria, algae and fungi in thermal reactors there by stopping the chances of biocorrosion and plugging of the pipe lines. (author). 5 refs., 1 fig

Reactor

International Nuclear Information System (INIS)

Toyama, Masahiro; Kasai, Shigeo.

1978-01-01

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

Heat Transfer Modeling of an Annular On-Line Spray Water Cooling Process for Electric-Resistance-Welded Steel Pipe.

Science.gov (United States)

Chen, Zejun; Han, Huiquan; Ren, Wei; Huang, Guangjie

2015-01-01

On-line spray water cooling (OSWC) of electric-resistance-welded (ERW) steel pipes can replace the conventional off-line heat treatment process and become an important and critical procedure. The OSWC process improves production efficiency, decreases costs, and enhances the mechanical properties of ERW steel pipe, especially the impact properties of the weld joint. In this paper, an annular OSWC process is investigated based on an experimental simulation platform that can obtain precise real-time measurements of the temperature of the pipe, the water pressure and flux, etc. The effects of the modes of annular spray water cooling and related cooling parameters on the mechanical properties of the pipe are investigated. The temperature evolutions of the inner and outer walls of the pipe are measured during the spray water cooling process, and the uniformity of mechanical properties along the circumferential and longitudinal directions is investigated. A heat transfer coefficient model of spray water cooling is developed based on measured temperature data in conjunction with simulation using the finite element method. Industrial tests prove the validity of the heat transfer model of a steel pipe undergoing spray water cooling. The research results can provide a basis for the industrial application of the OSWC process in the production of ERW steel pipes.

Heat Transfer Modeling of an Annular On-Line Spray Water Cooling Process for Electric-Resistance-Welded Steel Pipe

Science.gov (United States)

Chen, Zejun; Han, Huiquan; Ren, Wei; Huang, Guangjie

2015-01-01

On-line spray water cooling (OSWC) of electric-resistance-welded (ERW) steel pipes can replace the conventional off-line heat treatment process and become an important and critical procedure. The OSWC process improves production efficiency, decreases costs, and enhances the mechanical properties of ERW steel pipe, especially the impact properties of the weld joint. In this paper, an annular OSWC process is investigated based on an experimental simulation platform that can obtain precise real-time measurements of the temperature of the pipe, the water pressure and flux, etc. The effects of the modes of annular spray water cooling and related cooling parameters on the mechanical properties of the pipe are investigated. The temperature evolutions of the inner and outer walls of the pipe are measured during the spray water cooling process, and the uniformity of mechanical properties along the circumferential and longitudinal directions is investigated. A heat transfer coefficient model of spray water cooling is developed based on measured temperature data in conjunction with simulation using the finite element method. Industrial tests prove the validity of the heat transfer model of a steel pipe undergoing spray water cooling. The research results can provide a basis for the industrial application of the OSWC process in the production of ERW steel pipes. PMID:26201073

Applicability of ANSYS ELBOW290 element for flexibility calculation of tight radius bends on feeder pipes in CANDU reactors

Energy Technology Data Exchange (ETDEWEB)

Zhang, X., E-mail: Xuan.Zhang@candu.com [Candu Energy Inc, Mississauga, ON (Canada)

2015-07-01

A curved pipe element, ELBOW290, became available in ANSYS 12. This element was developed based on a simplified shell theory, and maintains the ability to capture cross-sectional deformations of elbows. Numerical testing on the applicability of this element for the flexibility calculation of the tight radius bends in CANDU reactors is carried out to determine the usability of this element in completing stress analyses for feeder pipes. Comparisons are made between the ELBOW290 and the shell element for various feeder bend types found in domestic and overseas CANDU reactors. The comparisons show that the ELBOW290 element is suitable for calculating the flexibility of the tight radius bends. (author)

Leak-before-break behaviour of nuclear piping systems

International Nuclear Information System (INIS)

Bartholome, G.; Wellein, R.

1992-01-01

The general concept for break preclusion of nuclear piping systems in the FRG consists of two main prerequisites: Basic safety; independent redundancies. The leak-before-break behaviour is open of these redundancies and will be verified by fracture mechanics. The following items have to be evaluated: The growth of detected and postulated defects must be negligible in one life time of the plant; the growth behaviour beyond design (i.e. multiple load collectives are taken into account) leads to a stable leak; This leakage of the piping must be detected by an adequate leak detection system long before the critical defect size is reached. The fracture mechanics calculations concerning growth and instability of the relevant defects and corresponding leakage areas are described in more detail. The leak-before-break behaviour is shown for two examples of nuclear piping systems in pressurized water reactors: main coolant line of SIEMENS-PWR 1300 MW (ferritic material, diameter 800 mm); surge line of Russian WWER 440 (austenitic material, diameter 250 mm). The main results are given taking into account the relevant leak detection possibilities. (author). 9 refs, 9 figs

Research Activities on Development of Piping Design Methodology of High Temperature Reactors

Energy Technology Data Exchange (ETDEWEB)

Huh, Nam-Su [Seoul National Univ. of Science and Technology, Seoul(Korea, Republic of); Won, Min-Gu [Sungkyukwan Univ., Suwon (Korea, Republic of); Oh, Young-Jin [KEPCO Engineering and Construction Co. Inc., Gimcheon (Korea, Republic of); Lee, Hyeog-Yeon; Kim, Yoo-Gon [Korea Atomic Energy Research Institute, Daejeon(Korea, Republic of)

2016-10-15

A SFR is operated at high temperature and low pressure compared with commercial pressurized water reactor (PWR), and such an operating condition leads to time-dependent damages such as creep rupture, excessive creep deformation, creep-fatigue interaction and creep crack growth. Thus, high temperature design and structural integrity assessment methodology should be developed considering such failure mechanisms. In terms of design of mechanical components of SFR, ASME B and PV Code, Sec. III, Div. 5 and RCC-MRx provide high temperature design and assessment procedures for nuclear structural components operated at high temperature, and a Leak-Before-Break (LBB) assessment procedure for high temperature piping is also provided in RCC-MRx, A16. Three web-based evaluation programs based on the current high temperature codes were developed for structural components of high temperature reactors. Moreover, for the detailed LBB analyses of high temperature piping, new engineering methods for predicting creep C*-integral and creep COD rate based either on GE/EPRI or on reference stress concepts were proposed. Finally, the numerical methods based on Garofalo's model and RCC-MRx have been developed, and they have been implemented into ABAQUS. The predictions based on both models were compared with the experimental results, and it has been revealed that the predictions from Garafalo's model gave somewhat successful results to describe the deformation behavior of Gr. 91 at elevated temperatures.

WWER type reactor primary loop imitation on large test loop facility in MARIA reactor

International Nuclear Information System (INIS)

Moldysh, A.; Strupchevski, A.; Kmetek, Eh.; Spasskov, V.P.; Shumskij, A.M.

1982-01-01

At present in Poland in cooperation with USSR a nuclear water loop test facility (WL) in 'MARIA' reactor in Sverke is under construction. The program objective is to investigate processes occuring in WWER reactor under emergency conditions, first of all after the break of the mainprimary loop circulation pipe-line. WL with the power of about 600 kW consists of three major parts: 1) an active loop, imitating the undamaged loops of the WWER reactor; 2) a passive loop assignedfor modelling the broken loop of the WWER reactor; 3) the emergency core cooling system imitating the corresponding full-scale system. The fuel rod bundle consists of 18 1 m long rods. They were fabricated according to the standard WWER fuel technology. In the report some general principles of WWERbehaviour imitation under emergency conditions are given. They are based on the operation experience obtained from 'SEMISCALE' and 'LOFT' test facilities in the USA. A description of separate modelling factors and criteria effects on the development of 'LOCA'-type accident is presented (the break cross-section to the primary loop volume ratio, the pressure differential between inlet and outlet reactor chambers, the pressure drop rate in the loop, the coolant flow rate throuh the core etc.). As an example a comparison of calculated flow rate variations for the WWER-1000 reactor and the model during the loss-of-coolant accident with the main pipe-line break at the core inlet is given. Calculations have been carried out with the use of TECH'-M code [ru

CSNI specialist meeting on leak-before-break in nuclear reactor piping: proceedings

Energy Technology Data Exchange (ETDEWEB)

1984-08-01

On September 1 and 2, 1983, the CSNI subcommittee on primary system integrity held a special meeting in Monterey, California, on the subject of leak-before-break in nuclear reactor piping systems. The purpose of the meeting was to provide an international forum for the exchange of ideas, positions, and research results; to identify areas requiring additional research and development; and to determine the general attitude toward acceptance of the leak-before-break concept. The importance of the leak-before-break issue was evidenced by excellent attendance at the meeting and through active participation by the meeting attendees. Approximately 125 people representing fifteen different nations attended the meeting. The meeting was divided into four technical sessions addressing the following areas: Application of Piping Fracture Mechanics to Leak-Before Break, Leak Rate and Leak Detection, Leak-Before-Break Studies, Methods and Results, Current and Proposed Positions on Leak-Before-Break.

CSNI specialist meeting on leak-before-break in nuclear reactor piping: proceedings

International Nuclear Information System (INIS)

1984-08-01

On September 1 and 2, 1983, the CSNI subcommittee on primary system integrity held a special meeting in Monterey, California, on the subject of leak-before-break in nuclear reactor piping systems. The purpose of the meeting was to provide an international forum for the exchange of ideas, positions, and research results; to identify areas requiring additional research and development; and to determine the general attitude toward acceptance of the leak-before-break concept. The importance of the leak-before-break issue was evidenced by excellent attendance at the meeting and through active participation by the meeting attendees. Approximately 125 people representing fifteen different nations attended the meeting. The meeting was divided into four technical sessions addressing the following areas: Application of Piping Fracture Mechanics to Leak-Before Break, Leak Rate and Leak Detection, Leak-Before-Break Studies, Methods and Results, Current and Proposed Positions on Leak-Before-Break

Development of FBR piping bellows joint

International Nuclear Information System (INIS)

Tsukimori, Kazuyuki; Iwata, Koji

1991-01-01

Reduction of construction cost is one of the most important problems to realize a FBR (Fast Breeder Reactor) Plant. Significant reduction of the construction cost of a reactor building, related equipments and facilities can be expected by shortening the length of its long cooling pipes. Since the bellows has a great capacity for absorbing thermal expansion displacement, application of bellows expansion joints is considered as the most influential measure for reduction of the piping length. To confirm technological possibilities of application and practical use of bellows joints in the main piping systems, extensive R and D's, development of various methods for evaluating the strength of bellows, establishment of inspection and maintenance techniques, studies on safety logic, etc., were carried out by PNC from 1983 to 1988. Through these studies, technological possibilities of bellows joints were confirmed and the results were summarized in the 'Structural Design Guide for Class 1 Piping Bellows Expansion Joints of Fast Breeder Reactor for Elevated Temperature Service' and the 'Inspection and Maintenance Standards of Piping bellows expansion Joints'. (author)

75 FR 42436 - Houston Pipe Line Company LP-Bammel Storage, Docket No. PR10-51-000, et. al.; Notice of Baseline...

Science.gov (United States)

2010-07-21

... DEPARTMENT OF ENERGY Federal Energy Regulatory Commission Houston Pipe Line Company LP--Bammel Storage, Docket No. PR10-51- 000, et. al.; Notice of Baseline Filings July 14, 2010. Houston Pipe Line..., 2010, respectively the applicants listed above submitted their baseline filing of its Statement of...

IEA-R1 renewed primary coolant piping system stress analysis

International Nuclear Information System (INIS)

Fainer, Gerson; Faloppa, Altair A.; Oliveira, Carlos A. de; Mattar Neto, Miguel

2015-01-01

A partial replacement of the IEA-R1 piping system was conducted in 2014. The aim of this work is to perform the stress analysis of the renewed primary piping system of the IEA-R1, taking into account the as built conditions and the pipe modifications. The nuclear research reactor IEA-R1 is a pool type reactor designed by Babcox-Willcox, which is operated by IPEN since 1957. The primary coolant system is responsible for removing the residual heat of the Reactor core. As a part of the life management, a regular inspection detected some degradation in the primary piping system. In consequence, part of the piping system was replaced. The partial renewing of the primary piping system did not imply in major piping layout modifications. However, the stress condition of the piping systems had to be reanalyzed. The structural stress analysis of the primary piping systems is now presented and the final results are discussed. (author)

75 FR 69052 - Certain Seamless Carbon and Alloy Steel Standard, Line, and Pressure Pipe From the People's...

Science.gov (United States)

2010-11-10

... threatened with material injury by reason of imports of seamless pipe from the PRC. According to section 736... that determination is based on the threat of material injury and is not accompanied by a finding that... Alloy Steel Standard, Line, and Pressure Pipe From the People's Republic of China: Amended Final...

New developments in velocity profile measurement and pipe wall wear monitoring for hydrotransport lines

Energy Technology Data Exchange (ETDEWEB)

O' Keefe, C.; Maron, R.J. [CiDRA Minerals Processing Inc., Wallingford, CT (United States); Fernald, M.; Bailey, T. [CiDRA Corporate Services, Wallingford, CT (United States); Van der Spek, A. [ZDOOR, Rotterdam (Netherlands)

2009-07-01

Sonar array flow measurement technology was initially developed a decade ago with the goal of non-invasively measuring multi-phase flows in the petroleum industry. The same technology was later adapted to the mineral processing industry where it has been rapidly adopted. The specific sensor technology, based on piezoelectric film sensors, provides unique measurement capabilities, including the ability to non-invasively measure localized strains in the walls of pipes. Combined with sonar array processing algorithms, an axial array of such sensors can measure flow velocities within a pipe. The sensors are useful for monitoring and managing slurry flow in horizontal pipes since they provide real-time velocity profiles measurement. The information is useful in determining the approach and onset of solid deposition on the bottom of the pipe. The sensors also provide a non-invasive measurement of pipe wear on slurry lines. Such measurements are currently made by hand-held portable ultrasonic thickness gages. The shortfalls associated with this manual method are overcome with a set of permanently or semi-permanently installed transducers clamped onto the outside of the pipe, where sensors measure the thickness of the pipe. This system and approach results in better repeatability and accuracy compared to manual methods. It also decreases inspection labor costs and pipe access requirements. It was concluded that the potential impact on personnel safety and environmental savings will be significant. 3 refs., 20 figs.

Gas cooled reactors

International Nuclear Information System (INIS)

Kojima, Masayuki.

1985-01-01

Purpose: To enable direct cooling of reactor cores thereby improving the cooling efficiency upon accidents. Constitution: A plurality sets of heat exchange pipe groups are disposed around the reactor core, which are connected by way of communication pipes with a feedwater recycling device comprising gas/liquid separation device, recycling pump, feedwater pump and emergency water tank. Upon occurrence of loss of primary coolants accidents, the heat exchange pipe groups directly absorb the heat from the reactor core through radiation and convection. Although the water in the heat exchange pipe groups are boiled to evaporate if the forcive circulation is interrupted by the loss of electric power source, water in the emergency tank is supplied due to the head to the heat exchange pipe groups to continue the cooling. Furthermore, since the heat exchange pipe groups surround the entire circumference of the reactor core, cooling is carried out uniformly without resulting deformation or stresses due to the thermal imbalance. (Sekiya, K.)

Device of connecting the metal sheet lining a concrete enclosure to a pipe opening inside the enclosure

International Nuclear Information System (INIS)

Petit, Guy.

1975-01-01

Said invention relates to a sealed device connecting a metal sheet anchored on the internal side of a concrete vessel containing a hot pressurized fluid, with a metallic pipe opening inside said vessel. It is intended for heat insulating structures so-called 'hot skin' used for the pressure vessels of some boiling water reactors. Said invention is intended for different types of said pipe such as: the penetrations for the inlets and outlets of the primary circuit, or anchoring cylindrical sheaths used as supports of components or other elements located inside said pressure vessel [fr

A simple approach to the prediction of waterhammer transients in a pipe line with entrapped air

International Nuclear Information System (INIS)

Epstein, Michael

2008-01-01

The pressure histories within entrapped air bubbles in a pipe line during a waterhammer transient are treated theoretically. A convenient integral method is introduced, which takes full account of air/water interface movement and liquid compressibility. The significance of the method is that it provides a simple equation set for approximating, with good accuracy and with a small degree of conservatism, the solution to a problem that otherwise involves coupled partial differential equations on time dependent domains with non-linear boundary conditions. The accuracy of the method is defined by its comparison with available numerical-solution-predictions and measurements of the pressure within an entrapped-air-bubble at a dead end in a pipe. The method is shown to be a computationally simple and efficient way of assessing the impact of liquid compressibility on pressure rise when multiple water columns and air pockets are present in a pipe line

Analysis of water hammer-structure interaction in piping system for a loss of coolant accident in primary loop of pressurized water reactor

International Nuclear Information System (INIS)

Zhang Xiwen; Yang Jinglong; He Feng; Wang Xuefang

2000-01-01

The conventional analysis of water hammer and dynamics response of structure in piping system is divided into two parts, and the interaction between them is neglected. The mechanism of fluid-structure interaction under the double-end break pipe in piping system is analyzed. Using the characteristics method, the numerical simulation of water hammer-structure interaction in piping system is completed based on 14 parameters and 14 partial differential equations of fluid-piping cell. The calculated results for a loss of coolant accident (LOCA) in primary loop of pressurized water reactor show that the waveform and values of pressure and force with time in piping system are different from that of non-interaction between water hammer and structure in piping system, and the former is less than the later

Reactor core

International Nuclear Information System (INIS)

Azekura, Kazuo; Kurihara, Kunitoshi.

1992-01-01

In a BWR type reactor, a great number of pipes (spectral shift pipes) are disposed in the reactor core. Moderators having a small moderating cross section (heavy water) are circulated in the spectral shift pipes to suppress the excess reactivity while increasing the conversion ratio at an initial stage of the operation cycle. After the intermediate stage of the operation cycle in which the reactor core reactivity is lowered, reactivity is increased by circulating moderators having a great moderating cross section (light water) to extend the taken up burnup degree. Further, neutron absorbers such as boron are mixed to the moderator in the spectral shift pipe to control the concentration thereof. With such a constitution, control rods and driving mechanisms are no more necessary, to simplify the structure of the reactor core. This can increase the fuel conversion ratio and control great excess reactivity. Accordingly, a nuclear reactor core of high conversion and high burnup degree can be attained. (I.N.)

Microstructure and Mechanical Properties of J55ERW Steel Pipe Processed by On-Line Spray Water Cooling

Directory of Open Access Journals (Sweden)

Zejun Chen

2017-04-01

Full Text Available An on-line spray water cooling (OSWC process for manufacturing electric resistance welded (ERW steel pipes is presented to enhance their mechanical properties and performances. This technique reduces the processing needed for the ERW pipe and overcomes the weakness of the conventional manufacturing technique. Industrial tests for J55 ERW steel pipe were carried out to validate the effectiveness of the OSWC process. The microstructure and mechanical properties of the J55 ERW steel pipe processed by the OSWC technology were investigated. The optimized OSWC technical parameters are presented based on the mechanical properties and impact the performance of steel pipes. The industrial tests show that the OSWC process can be used to efficiently control the microstructure, enhance mechanical properties, and improve production flexibility of steel pipes. The comprehensive mechanical properties of steel pipes processed by the OSWC are superior to those of other published J55 grade steels.

Acoustic emission for on-line reactor monitoring: results from field tests

International Nuclear Information System (INIS)

Hutton, P.H.; Kurtz, R.J.

1984-09-01

The objective of the acoustic emission (AE)/flaw characterization program is to develop use of the AE method on a continuous basis (during operation and during hydrotest) to detect and analyze flaw growth in reactor pressure vessels and primary piping. AE has the unique capability for continuous monitoring, high sensitivity, and remote flaw location

Residual-stresses in austenitic stainless-steel primary coolant pipes and welds of pressurized-water reactors

International Nuclear Information System (INIS)

Faure, F.; Leggatt, R.H.

1996-01-01

Surface and through thickness residual stress measurements were performed on an aged cast austenitic-ferritic stainless steel pipe and on an orbital TIG weld representative of those of primary coolant pipes in pressurized water reactors. An abrasive-jet hole drilling method and a block removal and layering method were used. Surface stresses and through thickness stress profiles are strongly dependent upon heat treatments, machining and welding operations. In the aged cast stainless steel pipe, stresses ranged between -250 and +175 MPa. On and near the orbital TIG weld, the outside surface of the weld was in tension both in the axial and hoop directions, with maximum values reaching 420 MPa in the weld. On the inside surface, the hoop stresses were compressive, reaching -300 MPa. However, the stresses in the axial direction at the root of the weld were tensile within 4 mm depth from the inside surface, locally reaching 280 MPa. (author)

Experimental study on air ingress during a primary pipe rupture accident with a graphite reactor core simulator

International Nuclear Information System (INIS)

Takeda, Tetsuaki; Hishida, Makoto; Baba, Shinichi

1991-11-01

When a primary coolant pipe of a High Temperature Gas Cooled Reactor (HTGR) ruptures, helium gas in the reactor core blows out into the container, and the primary cooling system reduces the pressure. After the pressures are balanced between the reactor and the container, air is expected to enter into the reactor core from the breach. It seems to be probable that the graphite structures is oxidized by air. Hence, it is necessary to investigate the air ingress process and the behavior of the generating gases by the oxidation reactions. The previous experimental study is performed on the molecular diffusion and natural convection of the two component gas mixtures using a test model simulating simply the reactor. Objective of the study was to investigate the air ingress process during the early stage of the primary pipe rupture accident. However, since the model did not have any kind of graphite components, the reaction between graphite and oxygen was not simulated. The present model includes the reactor core and the high temperature plenum simulators made of graphite. The major results obtained in the present study are summarized in the followings: (1) The air ingress process with graphite oxidation reaction is similar to that without the reaction qualitatively. (2) When the reactor core simulator is maintained at low temperatures (lower than 450degC), the initiation time of the natural circulation of air is almost equal to that of the natural circulation of nitrogen. On the other hand, when the temperature of the reactor core simulator is high (more than 500degC), the initiation time of the natural circulation of air is earlier than that of nitrogen. (3) When the temperature of the reactor core simulator is higher than 600degC, oxygen is almost dissipated by the graphite structures. When the temperature of the reactor core simulator is below 700degC, carbon dioxide mainly is generated by the oxidation reactions. (author)

Method for pre-heating lmfbr type reactors

International Nuclear Information System (INIS)

Yokozawa, Atsushi; Kataoka, Hajime.

1978-01-01

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

Stress-corrosion cracking in BWR and PWR piping

International Nuclear Information System (INIS)

Weeks, R.W.

1983-07-01

Intergranular stress-corrosion cracking of weld-sensitized wrought stainless steel piping has been an increasingly ubiquitous and expensive problem in boiling-water reactors over the last decade. In recent months, numerous cracks have been found, even in large-diameter lines. A number of potential remedies have been developed. These are directed at providing more resistant materials, reducing weld-induced stresses, or improving the water chemistry. The potential remedies are discussed, along with the capabilities of ultrasonic testing to find and size the cracks and related safety issues. The problem has been much less severe to date in pressurized-water reactors, reflecting the use of different materials and much lower coolant oxygen levels

Generation of cross section data of heat pipe working fluids for compact nuclear reactors

Energy Technology Data Exchange (ETDEWEB)

Slewinski, Anderson; Ribeiro, Guilherme B. [Instituto Tecnológico de Aeronáutica (ITA), São José dos Campos, SP (Brazil); Caldeira, Alexandre D., E-mail: anderson_sle@live.com, E-mail: alexdc@ieav.cta.br, E-mail: gbribeiro@ieav.cta.br [Instituto de Estudos Avançados (IEAv), São José dos Campos, SP (Brazil). Divisão de Energia Nuclear

2017-07-01

For compact nuclear power plants, such as the nuclear space propulsion proposed by the TERRA project, aspects like mass, size and efficiency are essential drivers that must be managed during the project development. Moreover, for high temperature reactors, the use of liquid metal heat pipes as the heat removal mechanism provides some important advantages as simplicity and reliability. Considering these aforementioned aspects, this paper aims the development of the procedure necessary to calculate the microscopic absorption cross section data of several liquid metal to be used as working fluids with heat pipes; which will be later compared with the given data from JEF Report ⧣14. The information necessary to calculate the cross section data will be obtained from the latest ENDF library version. The NJOY system will be employed with the following modules: RECONR, BROADR, UNRESR and GROUPR, using the same specifications used to calculate the cross section data encountered in the JEF Report ⧣14. This methodology allows a comparison with published values, verifying the procedure developed to calculate the microscopic absorption cross section for selected isotopes using the TERRA reactor spectrum. Liquid metals isotopes of Sodium (Na), Lithium (Li), Thallium (TI) and Mercury (Hg) are part of this study. (author)

Reactor water sampling device

International Nuclear Information System (INIS)

Sakamaki, Kazuo.

1992-01-01

The present invention concerns a reactor water sampling device for sampling reactor water in an in-core monitor (neutron measuring tube) housing in a BWR type reactor. The upper end portion of a drain pipe of the reactor water sampling device is attached detachably to an in-core monitor flange. A push-up rod is inserted in the drain pipe vertically movably. A sampling vessel and a vacuum pump are connected to the lower end of the drain pipe. A vacuum pump is operated to depressurize the inside of the device and move the push-up rod upwardly. Reactor water in the in-core monitor housing flows between the drain pipe and the push-up rod and flows into the sampling vessel. With such a constitution, reactor water in the in-core monitor housing can be sampled rapidly with neither opening the lid of the reactor pressure vessel nor being in contact with air. Accordingly, operator's exposure dose can be reduced. (I.N.)

Metallurgical factors that contribute to cracking in BWR piping

International Nuclear Information System (INIS)

Weeks, J.R.

1975-01-01

During the fall of 1974 and early winter of 1975, cracks have been discovered in the 4 in. bypass lines of several Boiling Water Reactors (BWR's) in the United States. Further, similar cracks were discovered at two BWR's in Japan during the same period. More recently, cracks have been discovered in the core spray piping and in a furnace-sensitized ''safe end'' and adjacent ''dutchman'' at the Dresden Nuclear Power Station, Unit No. 2. Although inspections at all other U.S. BWR's have not disclosed further instances of cracking in core spray piping, leaking cracks have been found in the core spray piping of two BWR's overseas. Metallurgical examinations of these cracks are not yet complete. The following observations have been made to date. All cracks (except those in the furnace-sensitized safe end and dutchman) occurred in seamless type 304 stainless steel piping or in elbows fabricated from such piping, in the outer heat affected zone of either field or shop welds, in lines isolated from the main primary coolant flow during full power operation, except for the not yet examined cracks in the Monticello bypass lines. The cracks are exclusively intergranular, and occur in metal that has been lightly sensitized by the welding process, with only intermittent grain boundary carbides. They developed in the areas of peak axial residual stresses from welding rather than in the most heavily sensitized areas. No fatigue striations have been found on the fracture surfaces. The evidence received to date strongly indicates that these cracks were caused by intergranular stress corrosion of weld-sensitized stainless steel by BWR water containing greater than 0.2 ppM oxygen. The possible role of fatigue or alternating stresses in this corrosion is not clear. Further, not all the cracks detected to date necessarily have occurred by the same mechanism

Effects of phosphate addition on biofilm bacterial communities and water quality in annular reactors equipped with stainless steel and ductile cast iron pipes.

Science.gov (United States)

Jang, Hyun-Jung; Choi, Young-June; Ro, Hee-Myong; Ka, Jong-Ok

2012-02-01

The impact of orthophosphate addition on biofilm formation and water quality was studied in corrosion-resistant stainless steel (STS) pipe and corrosion-susceptible ductile cast iron (DCI) pipe using cultivation and culture-independent approaches. Sample coupons of DCI pipe and STS pipe were installed in annular reactors, which were operated for 9 months under hydraulic conditions similar to a domestic plumbing system. Addition of 5 mg/L of phosphate to the plumbing systems, under low residual chlorine conditions, promoted a more significant growth of biofilm and led to a greater rate reduction of disinfection by-products in DCI pipe than in STS pipe. While the level of THMs (trihalomethanes) increased under conditions of low biofilm concentration, the levels of HAAs (halo acetic acids) and CH (chloral hydrate) decreased in all cases in proportion to the amount of biofilm. It was also observed that chloroform, the main species of THM, was not readily decomposed biologically and decomposition was not proportional to the biofilm concentration; however, it was easily biodegraded after the addition of phosphate. Analysis of the 16S rDNA sequences of 102 biofilm isolates revealed that Proteobacteria (50%) was the most frequently detected phylum, followed by Firmicutes (10%) and Actinobacteria (2%), with 37% of the bacteria unclassified. Bradyrhizobium was the dominant genus on corroded DCI pipe, while Sphingomonas was predominant on non-corroded STS pipe. Methylobacterium and Afipia were detected only in the reactor without added phosphate. PCR-DGGE analysis showed that the diversity of species in biofilm tended to increase when phosphate was added regardless of the pipe material, indicating that phosphate addition upset the biological stability in the plumbing systems.

Nuclear reactor

International Nuclear Information System (INIS)

Miyashita, Akio.

1981-01-01

Purpose: To facilitate and accelerate a leakage test of valves of a main steam pipe by adding a leakage test partition valve thereto. Constitution: A leakage testing partition valve is provided between a pressure vessel for a nuclear reactor and the most upstream side valve of a plurality of valves to be tested for leakage, a testing branch pipe is communicated with the downstream side of the partition valve, and the testing water for preventing leakage is introduced thereto through the branch pipe. Since main steam pipe can be simply isolated by closing the partition valve in the leakage test, the leakage test can be conducted without raising or lowering the water level in the pressure vessel, and since interference with other work in the reactor can be eliminated, the leakage test can be readily conducted parallel with other work in the reactor in a short time. Clean water can be used without using reactor water as the test water. (Yoshihara, H.)

Vibration monitoring of the primary piping systems during the hot functional tests of the Mulheim-Karlich PWR

International Nuclear Information System (INIS)

Bauernfeind, V.; Bloem, T.; Pache, W.; Diederich, H.J.

1989-01-01

During the hot functional tests of the Muelheim--Kaerlich first-of-a-kind plant, vibration measurements were made on the reactor pressure vessel and its' internals and on the primary piping system and main coolant pumps. This paper contains results of the measurements taken on the pipes and the pumps with an interpretation of these measurements based on an analytical model of the primary system. The main aim of the measurement program is to confirm that the components, which are of new design, are adequately dimensioned for the operational vibration loads during the service life of the reactor. In addition, the vibrational modes of the hot lines, the steam generators and the pumps with the adjacent cold lines were determined. These values were compared with the analytically calculated resonance frequencies and eigenforms. Good agreement was found. In the course of these comparisons, information on the modelling of the supporting structures and the efficiency of the damping elements during normal operation was obtained

Resolution of concerns in auxiliary feedwater piping

International Nuclear Information System (INIS)

Bain, R.A.; Testa, M.F.

1994-01-01

Auxiliary feedwater piping systems at pressurized water reactor (PWR) nuclear power plants have experienced unanticipated operating conditions during plant operation. These unanticipated conditions have included plant events involving backleakage through check valves, temperatures in portions of the auxiliary feedwater piping system that exceed design conditions, and the occurrence of unanticipated severe fluid transients. The impact of these events has had an adverse effect at some nuclear stations on plant operation, installed plant components and hardware, and design basis calculations. Beaver Valley Unit 2, a three loop pressurized water reactor nuclear plant, has observed anomalies with the auxiliary feedwater system since the unit went operational in 1987. The consequences of these anomalies and plant events have been addressed and resolved for Beaver Valley Unit 2 by performing engineering and construction activities. These activities included pipe stress, pipe support and pipe rupture analysis, the monitoring of auxiliary feedwater system temperature and pressure, and the modification to plant piping, supports, valves, structures and operating procedures

Development of an ISI robot for the fast breeder reactor MONJU primary heat transfer system piping

International Nuclear Information System (INIS)

Tagawa, Akihiro; Ueda, Masashi; Yamashita, Takuya; Narisawa, Masataka; Haga, Kouichi

2011-01-01

This paper describes the development of a new inspection robot for the In-Service Inspection of the heat transfer system of the Fast Breeder Reactor MONJU. The inspection was carried out using a tire-type ultrasonic sensor for volumetric tests at high temperature (atmosphere, 55degC; piping surface, 80degC) and radiation exposure condition (dose rate, 10 mGy/h; piping surface dose rate, 15 mGy/h). An inspection robot using a new tire type for the ultrasonic testing sensor and a new control method was developed. A signal-to-noise ratio S/N over 2 was obtained during the functional test for a calibration defect with a depth of 50%t (from the tube wall thickness). In the automatic inspection test, an EDM slit with a depth of 9% from the pipe thickness was detectable and with an S/N ratio = 4.0 (12.0 dB). (author)

Effects of blast wave to main steam piping under high energy line break condition by TNT model

Energy Technology Data Exchange (ETDEWEB)

Kim, Seung Hyun; Lee, Eung Seok; Chang, Yoon Suk [Kyung Hee University, Yongin (Korea, Republic of)

2016-05-15

The aim of this study is to examine effect of the blast wave according to pipe break position through FE (Finite Element) analyses. If HELB (High Energy Line Break) accident occurs in nuclear power plants, not only environmental effect such as release of radioactive material but also secondary structural defects should be considered. Sudden pipe rupture causes ejection of high temperature and pressure fluid, which acts as a blast wave around the break location. The blast wave caused by the HELB has a possibility to induce structural defects around the components such as safe-related injection pipes and other structures.

Elastic-plastic response of a piping system due to simulated double-ended guillotine break events

International Nuclear Information System (INIS)

Kussmaul, K.; Diem, H.; Hunger, H.; Katzenmeier, G.

1987-01-01

From the blowdown experiments performed on the HDR feedwater line with feedwater check valve the conclusion can be drawn that high transient loads of up to plastic strains of 3%, acting on an initially integer piping system, can be sustained without loss of integrity for a low number of load cycles due to the plasticizing capacity of the pipework materials nowadays used in reactor technology. In the experiments carried out with ferritic piping of ND 400 pressure peaks up to about 31,5 MPa were achieved which resulted in excessive strains of up to 3%. By nonlinear finite element computations (ABAQUS) it was possible to describe the elastic-plastic behaviour of the piping in a good approximation. (orig./GL)

Gamma-radiography techniques applied to quality control of welds in water pipe lines

International Nuclear Information System (INIS)

Sanchez, W.; Oki, H.

1974-01-01

Non-destructive testing of welds may be done by the gamma-radiography technique, in order to detect the presence or absence of discontinuities and defects in the bulk of deposited metal and near the base metal. Gamma-radiography allows the documentation of the test with a complete inspection record, which is a fact not common in other non-destructive testing methods. In the quality control of longitudinal or transversal welds in water pipe lines, two exposition techniques are used: double wall and panoramic exposition. Three different water pipe lines systems have analysed for weld defects, giving a total of 16,000 gamma-radiographies. The tests were made according to the criteria established by the ASME standard. The principal metallic discontinuites found in the weld were: porosity (32%), lack of penetration (29%), lack of fusion (20%), and slag inclusion (19%). The percentage of gamma-radiographies showing welds without defects was 39% (6168 gamma-radiographies). On the other hand, 53% (8502 gamma-radiographies) showed the presence of acceptable discontinuities and 8% (1330 gamma-radiographies) were rejected according to the ASME standards [pt

Analysis of gamma ray intensity on the S/C vent pipes area in the unit 2 reactor building of the Fukushima Daiichi Nuclear Power Plant

Energy Technology Data Exchange (ETDEWEB)

Cho, Jai Wan; Choi, Young Soo; Jeong, Kyung Min [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2014-05-15

The robot is equipped with cameras, a dosimeter, and 2 DOF (degree of freedom) manipulation arms. It loads a small vehicle equipped with a camera that can access and inspect narrow areas. TEPCO is using the four-legged walking robot to inspect the suppression chamber (S/C) area of the unit 2 reactor building basement in the Fukushima Daiichi Nuclear Power Plant. The robot carried out 6 missions for about four months, from 11 December, 2012 to 15 March, 2013, where it examined an evidence of a leakage of radioactivity contaminated water in the S/C area of unit 2 reactor building. When a camera's signal processing unit, which is consist of ASIC and FPGA devices manufactured by a CMOS fabrication process, is exposed to a higher dose rate gamma ray, the speckle distribution in the camera image increase more. From the inspection videos, released by TEPCO, of the underground 8 vent pipes in the unit 2 reactor building, we analyzed the speckle distribution from the high dose-rate gamma rays. Based on the distribution of the speckle, we attempted to characterize the vent pipe with much radioactivity contaminated materials among the eight vent pipes connected to the PCV. The numbers of speckles viewed in the image of a CCD (or CMOS) camera are related to an intensity of the gamma ray energy emitted by a nuclear fission reaction from radioactivity materials. The numbers of speckles generated by gamma ray irradiation in the camera image are calculated by an image processing technique. Therefore, calculating the speckles counts, we can determine the vent pipe with relatively most radioactivity-contaminated materials among the other vent pipes. From the comparison of speckles counts calculated in the inspection image of the vent pipe with the speckles counts extracted by gamma ray irradiation experiment of the same small vehicle camera model loaded with the four-legged walking robot, we can qualitatively estimate the gamma ray dose-rate in the S/C vent pipe area of the

Solid-Core Heat-Pipe Nuclear Batterly Type Reactor

International Nuclear Information System (INIS)

Ehud Greenspan

2008-01-01

This project was devoted to a preliminary assessment of the feasibility of designing an Encapsulated Nuclear Heat Source (ENHS) reactor to have a solid core from which heat is removed by liquid-metal heat pipes (HP). Like the SAFE 400 space nuclear reactor core, the HPENHS core is comprised of fuel rods and HPs embedded in a solid structure arranged in a hexagonal lattice in a 3:1 ratio. The core is oriented horizontally and has a square rather cylindrical cross section for effective heat transfer. The HPs extend from the two axial reflectors in which the fission gas plena are embedded and transfer heat to an intermediate coolant that flows by natural-circulation. The HP-ENHS is designed to preserve many features of the ENHS including 20-year operation without refueling, very small excess reactivity throughout life, natural circulation cooling, walkaway passive safety, and robust proliferation resistance. The target power level and specific power of the HP-ENHS reactor are those of the reference ENHS reactor. Compared to previous ENHS reactor designs utilizing a lead or lead-bismuth alloy natural circulation cooling system, the HP-ENHS reactor offers a number of advantageous features including: (1) significantly enhanced passive decay heat removal capability; (2) no positive void reactivity coefficients; (3) relatively lower corrosion of the cladding (4) a core that is more robust for transportation; (5) higher temperature potentially offering higher efficiency and hydrogen production capability. This preliminary study focuses on five areas: material compatibility analysis, HP performance analysis, neutronic analysis, thermal-hydraulic analysis and safety analysis. Of the four high-temperature structural materials evaluated, Mo TZM alloy is the preferred choice; its upper estimated feasible operating temperature is 1350 K. HP performance is evaluated as a function of working fluid type, operating temperature, wick design and HP diameter and length. Sodium is the

Rectification of leak from upper aluminium thermal shield cooling water inlet line of Cirus reactor

International Nuclear Information System (INIS)

Bhatnagar, Anil; Joshi, N.S.; Kharpate, A.V.; Marik, S.K.

2006-01-01

During 1994, a small water leak was observed from the upper aluminium thermal shield of Cirus reactor. Detailed investigations revealed that the leakage was from the weld joint of one of the 1 1/4 inch NB Sch. 80 coolant inlet pipes connected to the upper aluminium thermal shield. The location of the leak was identified by monitoring the stabilised water level in the vertical inlet pipe under stagnant condition. The exact location was identified by installing an inflatable seal arrangement inside the leaky pipe and inflating the seal at different elevations to isolate the leaky location and ensuring that the leak was completely stopped. This location was about 15 feet below the operating floor of the reactor. The pipe was visually inspected with the help of a fibre-scope to assess the condition of the inner surface. Eddy current testing was also carried out for volumetric examination. This revealed one more localised flaw on the outer surface little above the leaky joint. A hollow plug, with expandable rings, having C-shaped cross section at both the ends and a straight portion in the middle to cover the defective region, was developed and qualified in a mock-up station after extensive trials. In view of the site constraints, a flexible hollow link assembly was engineered, for installing the plug remotely. The inner surface of the pipe was cleaned using an emery brush and a deburring tool. The plug was then installed covering the leak area and the rings were expanded by remote tightening. The shield was hydro-tested satisfactorily. (author)

Characterization of bond line discontinuities in a high-Mn TWIP steel pipe welded by HF-ERW

Energy Technology Data Exchange (ETDEWEB)

Park, Gitae; Kim, Bongyoon; Kang, Yongjoon [Division of Materials Science and Engineering, Hanyang University, 222, Wangsimni-ro, Seongdong-gu, Seoul 04763 (Korea, Republic of); Kang, Heewoong [RD Team, Husteel, 131 Bugokgongdan-ro, Songak-eup, Dangjin-si, Chungnam 31721 (Korea, Republic of); Lee, Changhee, E-mail: chlee@hanyang.ac.kr [Division of Materials Science and Engineering, Hanyang University, 222, Wangsimni-ro, Seongdong-gu, Seoul 04763 (Korea, Republic of)

2016-08-15

In this work, the microstructure and defects in a high-frequency electrical resistance welded (HF-ERW) pipe of high-Mn twinning-induced plasticity (TWIP) steel were characterized. The microstructure of the base metal and the bond line were examined using both optical microscopy and scanning electron microscopy. The features of the bond line were similar to those of conventional steel. Simultaneously, the circumferential ductility was evaluated via a flaring test. It was concluded that the deterioration of the circumferential ductility in a high-Mn TWIP steel pipe was caused by irregular shaped oxide defects and a penetrator that had been formed during welding. Specifically, the penetrator, which is composed of MnO and Mn{sub 2}SiO{sub 4}, was found to be the most influential on the circumferential ductility of the welded pipe. The penetrator was analyzed using both an electron probe micro analyzer and transmission electron microscopy, and the formation sequence of the penetrator was evaluated. - Highlights: •This study focused on applying the HF-ERW process to the seam welding of expandable pipe using TWIP steels. •For improvement of the circumferential ductility, deterioration factors were characterized. •Penetrator which would mainly deteriorate the circumferential ductility consisted of round MnO and Mn{sub 2}SiO{sub 4}. •Metallurgical evidence of existing theory regarding the mechanism of defect formation during the HF-ERW was characterized.

49 CFR 195.424 - Pipe movement.

Science.gov (United States)

2010-10-01

... 49 Transportation 3 2010-10-01 2010-10-01 false Pipe movement. 195.424 Section 195.424... PIPELINE Operation and Maintenance § 195.424 Pipe movement. (a) No operator may move any line pipe, unless... in the line section involved are joined by welding unless— (1) Movement when the pipeline does not...

Design features of BREST reactors. Experimental work to advance the concept of BREST reactors. Results and plans

International Nuclear Information System (INIS)

Filin, A.I.; Orlov, V.V.; Leonov, V.N.; Sila-Novitskij, A.G.; Smirnov, V.S.; Tsikunov, V.S.

2001-01-01

Principle designs of 300 MW(th) and 1200 MW(th) lead-cooled fast reactors are presented. Reactors of various output are shown to be built using the same principles. In conjunction with increased output and to implement inherent safety concept in BREST-1200 reactor design a number of new solutions, which may be used in BREST-300 concept too, has been taken including: pool-type reactor design not requiring metal vessel, hence, not limiting reactor power; new handling system allowing to reduce central hall and building dimensions as a whole; emergency cooling system using Field pipes, immersed directly in lead, which may be used to cool down reactor under normal conditions; by-pass line incorporated in coolant loop allowing to refuse the actively actuating valve initiated in pumps shut down. (author)

The influence of prefabricated pipe cement coatings and those made during pipe renovation on drinking water quality

OpenAIRE

Młyńska Anna; Zielina Michał

2017-01-01

Nowadays, cement coatings are often used as an anticorrosion protection of the internal surfaces of manufactured ductile iron water pipes. The protective cement linings are also commonly used for old water pipe renovation. In both cases, the cement lining is an excellent anticorrosion protection of the pipelines, effectively separating the pipe wall from the flowing water. Moreover, cement linings protect the pipelines not only by a mechanical barrier, but also by a chemical barrier creating ...

Careful determination of inservice inspection of piping by computer analysis in nuclear power plant

International Nuclear Information System (INIS)

Lim, H. T.; Lee, S. L.; Lee, J. P.; Kim, B. C.

1992-01-01

Stress analysis has been performed using computer program ANSYS in the pressurizer surge line in order to predict possibility of crack generation due to thermal stratification phenomena in pipes connected to reactor coolant system of Nuclear power plants. Highly vulnerable area to crack generation has been chosen by the analysis of fatigue due to thermal stress in pressurizer surge line. This kind of result can be helpful to choose the location requiring intensive care during inservice inspection of nuclear power plants.

Modifications and modernization of the Portuguese research reactor (RPI)

International Nuclear Information System (INIS)

Cardeira, F.M.; Menezes, J.B.

1995-01-01

The Portuguese Research Reactor (RPI) reached its criticality in April 1961 and has successfully operated for more than 30 years without important incidents. Several replacements of equipment and improvements were introduced during this period, the most important occurring in the modernisation period (1987-1991), with the purpose of improving safety and reliability of the reactor exploitation. The reactor has been shut-down during more than two years for important works of replacement and refurbishment of the primary piping and pool lining. The objective of this paper is to describe the main works performed on RPI reactor during its life time concerning replacements, upgrading and modernisation of reactor equipment and installations. (orig.)

Preliminary feasibility study of the heat - pipe ENHS reactor

International Nuclear Information System (INIS)

Fratoni, M.; Kim, L.; Mattafirri, S.; Petroski, R.; Greenspan, E.

2007-01-01

This preliminary study assesses the feasibility of designing an Encapsulated Nuclear Heat Source (ENHS) reactor [1] to have a solid core from which heat is removed by liquid-metal heat pipes (HP). Like the SAFE space nuclear reactor core [2], the HP-ENHS core is comprised of fuel rods and HPs embedded in a solid structure arranged in a hexagonal lattice in a 3:1 ratio. The HPs extend beyond the core length and transfer heat to a secondary coolant that flows by natural circulation. The HP-ENHS reactor is designed to preserve many features of the ENHS reactor including 20-year operation without refueling, very small excess reactivity throughout life, natural circulation cooling, walk-away passive safety, and robust proliferation resistance. The target power level and specific power of the HP-ENHS reactor are those of the reference ENHS reactor [1]. Compared to previous ENHS reactor designs utilizing a lead or lead-bismuth alloy natural circulation cooling system, the HP-ENHS reactor offers a number of possible advantageous features including: (1) significantly enhanced decay heat removal capability; (2) no positive void reactivity coefficients; (3) no direct contact between the fuel clad and coolant, hence, relatively lower wet corrosion of the clad; (4) a core that is more robust for transportation; (5) higher temperature potentially offering higher efficiency and hydrogen production capability. The study focuses on four areas: material compatibility analysis, HP performance analysis, neutronic analysis and thermal-hydraulic analysis. Of four high-temperature structural materials evaluated, Mo TZM alloy is the preferred choice; its upper estimated feasible operating temperature is 1350 K. HP performance is evaluated as a function of working fluid type, operating temperature, wick design and HP diameter and length. Sodium is the preferred working fluid and the HP working temperature is 1300 K. The neutronic analysis found that it is possible to achieve criticality

Probabilistic based design rules for intersystem LOCAS in ABWR piping

International Nuclear Information System (INIS)

Ware, A.G.; Wesley, D.A.

1993-01-01

A methodology has been developed for probability-based standards for low-pressure piping systems that are attached to the reactor coolant loops of advanced light water reactors (ALWRs) which could experience reactor coolant loop temperatures and pressures because of multiple isolation valve failures. This accident condition is called an intersystem loss-of-coolant accident (ISLOCA). The methodology was applied to various sizes of carbon and stainless steel piping designed to advanced boiling water reactor (ABWR) temperatures and pressures

Hot Leg Piping Materials Issues

International Nuclear Information System (INIS)

V. Munne

2006-01-01

With Naval Reactors (NR) approval of the Naval Reactors Prime Contractor Team (NRPCT) recommendation to develop a gas cooled reactor directly coupled to a Brayton power conversion system as the space nuclear power plant (SNPP) for Project Prometheus (References a and b) the reactor outlet piping was recognized to require a design that utilizes internal insulation (Reference c). The initial pipe design suggested ceramic fiber blanket as the insulation material based on requirements associated with service temperature capability within the expected range, very low thermal conductivity, and low density. Nevertheless, it was not considered to be well suited for internal insulation use because its very high surface area and proclivity for holding adsorbed gases, especially water, would make outgassing a source of contaminant gases in the He-Xe working fluid. Additionally, ceramic fiber blanket insulating materials become very friable after relatively short service periods at working temperatures and small pieces of fiber could be dislodged and contaminate the system. Consequently, alternative insulation materials were sought that would have comparable thermal properties and density but superior structural integrity and greatly reduced outgassing. This letter provides technical information regarding insulation and materials issues for the Hot Leg Piping preconceptual design developed for the Project Prometheus space nuclear power plant (SNPP)

Leak detection in the primary reactor coolant piping of nuclear power plant by applying beam-microphone technology

International Nuclear Information System (INIS)

Kasai, Yoshimitsu; Shimanskiy, Sergey; Naoi, Yosuke; Kanazawa, Junichi

2004-01-01

A microphone leak detection method was applied to the inlet piping of the ATR-prototype reactor, Fugen. Statistical analysis results showed that the cross-correlation method provided the effective results for detection of a small leakage. However, such a technique has limited application due to significant distortion of the signals on the reactor site. As one of the alternative methods, the beam-microphone provides necessary spatial selectivity and its performance is less affected by signal distortion. A prototype of the beam-microphone was developed and then tested at the O-arai Engineering Center of the Japan Nuclear Cycle Development Institute (JNC). On-site testing of the beam-microphone was carried out in the inlet piping room of an RBMK reactor of the Leningrad Nuclear Power Plant (LNPP) in Russia. A leak sound imitator was used to simulate the leakage sound under the leakage flow condition of 1-3 gpm (0.23-0.7 m 3 /h). Analysis showed that signal distortion does not seriously affect the performance of this method, and that sound reflection may result in the appearance of ghost sound sources. The test results showed that the influences of sound reflection and background noise were smaller at the high frequencies where the leakage location could be estimated with an angular accuracy of 5deg which is the range of localization accuracy required for the leak detection system. (author)

Peach Bottom Atomic Power Station recirc pipe dose rates with zinc injection and condenser replacement

International Nuclear Information System (INIS)

DiCello, D.C.; Odell, A.D.; Jackson, T.J.

1995-01-01

Peach Bottom Atomic Power Station (PBAPS) is located near the town of Delta, Pennsylvania, on the west bank of the Susquehanna River. It is situated approximately 20 miles south of Lancaster, Pennsylvania. The site contains two boiling water reactors of General Electric design and each rated at 3,293 megawatts thermal. The units are BWR 4s and went commercial in 1977. There is also a decommissioned high temperature gas-cooled reactor on site, Unit 1. PBAPS Unit 2 recirc pipe was replaced in 1985 and Unit 3 recirc pipes replaced in 1988 with 326 NGSS. The Unit 2 replacement pipe was electropolished, and the Unit 3 pipe was electropolished and passivated. The Unit 2 brass condenser was replaced with a Titanium condenser in the first quarter of 1991, and the Unit 3 condenser was replaced in the fourth quarter of 1991. The admiralty brass condensers were the source of natural zinc in both units. Zinc injection was initiated in Unit 2 in May 1991, and in Unit 3 in May 1992. Contact dose rate measurements were made in standard locations on the 28-inch recirc suction and discharge lines to determine the effectiveness of zinc injection and to monitor radiation build-up in the pipe. Additionally, HPGe gamma scans were performed to determine the isotopic composition of the oxide layer inside the pipe. In particular, the specific (μCi/cm 2 ) of Co-60 and Zn-65 were analyzed

Peach Bottom Atomic Power Station recirc pipe dose rates with zinc injection and condenser replacement

Energy Technology Data Exchange (ETDEWEB)

DiCello, D.C.; Odell, A.D.; Jackson, T.J. [PECO Energy Co., Delta, PA (United States)

1995-03-01

Peach Bottom Atomic Power Station (PBAPS) is located near the town of Delta, Pennsylvania, on the west bank of the Susquehanna River. It is situated approximately 20 miles south of Lancaster, Pennsylvania. The site contains two boiling water reactors of General Electric design and each rated at 3,293 megawatts thermal. The units are BWR 4s and went commercial in 1977. There is also a decommissioned high temperature gas-cooled reactor on site, Unit 1. PBAPS Unit 2 recirc pipe was replaced in 1985 and Unit 3 recirc pipes replaced in 1988 with 326 NGSS. The Unit 2 replacement pipe was electropolished, and the Unit 3 pipe was electropolished and passivated. The Unit 2 brass condenser was replaced with a Titanium condenser in the first quarter of 1991, and the Unit 3 condenser was replaced in the fourth quarter of 1991. The admiralty brass condensers were the source of natural zinc in both units. Zinc injection was initiated in Unit 2 in May 1991, and in Unit 3 in May 1992. Contact dose rate measurements were made in standard locations on the 28-inch recirc suction and discharge lines to determine the effectiveness of zinc injection and to monitor radiation build-up in the pipe. Additionally, HPGe gamma scans were performed to determine the isotopic composition of the oxide layer inside the pipe. In particular, the specific ({mu}Ci/cm{sup 2}) of Co-60 and Zn-65 were analyzed.

Refurbishment of the IEAR1 primary coolant system piping supports

International Nuclear Information System (INIS)

Fainer, Gerson; Faloppa, Altair A.; Oliveira, Carlos A. de; Mattar Neto, Miguel

2015-01-01

A partial replacement of the IEA-R1 piping system was concluded in 2014. This paper presents the study and the structural analysis of the IEA-R1 primary circuit piping supports, considering all the changes involved in the replacement. The IEA-R1 is a nuclear reactor for research purposes designed by Babcox-Willcox that is operated by IPEN since 1957. The reactor life management and modernization program is being conducted for the last two decades and already resulted in a series of changes, especially on the reactor coolant system. This set of components, divided in primary and secondary circuit, is responsible for the circulation of water into the core to remove heat. In the ageing management program that includes regular inspection, some degradation was observed in the primary piping system. As result, the renewing of the piping system was conducted in 2014. Moreover the poor condition of some original piping supports gave rise to the refurbishment of all piping supports. The aim of the present work is to review the design of the primary system piping supports taking into account the current conditions after the changes and refurbishment. (author)

Pressure tube type research reactor

International Nuclear Information System (INIS)

Ueda, Hiroshi.

1975-01-01

Object: To permit safe and reliable replacement of primary pipes by providing a reactor container so as to surround a pressure pipe, with upper portions of the two separably coupled together, and coupling the pressure pipe and primary piping by joint coupling above and below the reactor container, with the lower coupling joint surrounded by drain receptacle. Structure: At the time of replacement of a pressure pipe, a partition valve is opened to exhaust primary cooling water within pressure pipe and upper and lower portions of the primary piping and replace the decelerator within the reactor container with water of the same quality as that of pool water within an upper shield pool. Thereafter, the entire space above the drain receptacle is filled with pool water by closing a partition valve and opening a water supply valve. Then, upper portion seal cover, pool bottom lid, upper joint and upper portion primary piping are removed, then bolts and nuts are loosened, and the pressure pipe is taken out together with the shield block. (Kamimura, M.)

Study of verification and validation of standard welding procedure specifications guidelines for API 5L X-70 grade line pipe welding

Directory of Open Access Journals (Sweden)

Qazi H. A. A.

2017-12-01

Full Text Available Verification and validation of welding procedure specifications for X-70 grade line pipe welding was performed as per clause 8.2, Annexure B and D of API 5L, 45th Edition to check weld integrity in its future application conditions. Hot rolled coils were imported from China, de-coiling, strip edge milling, three roller bending to from pipe, inside and outside submerged arc welding of pipe, online ultrasonic testing of weld, HAZ and pipe body, cutting at fixed random length of pipe, visual inspection of pipe, Fluoroscopic inspection of pipe, welding procedure qualification test pieces marking at weld portion of the pipe, tensile testing, guided bend testing, CVN Impact testing were performed. Detailed study was conducted to explore possible explanations and variation in mechanical properties, WPS is examined and qualified as per API 5L 45th Edition.

An in-pipe mobile micromachine using fluid power. A mechanism adaptable to pipe diameters

International Nuclear Information System (INIS)

Yoshida, Kazuhiro; Yokota, Shinichi; Takahashi, Ken

2000-01-01

To realize micro maintenance robots for small diameter pipes of nuclear reactors and so on, high power in-pipe mobile micromachines have been required. The authors have proposed the bellows microactuator using fluid power and have tried to apply the actuators to in-pipe mobile micromachines. In the previous papers, some inchworm mobile machine prototypes with 25 mm in diameter are fabricated and the traveling performances are experimentally investigated. In this paper, to miniaturize the in-pipe mobile machine and to make it adaptable to pipe diameters, firstly, a simple rubber-tube actuator constrained with a coil-spring is proposed and the static characteristics are investigated. Secondly, a supporting mechanism which utilizes a toggle mechanism and is adaptable to pipe diameters is proposed and the supporting forces are investigated. Finally, an in-pipe mobile micromachine for pipe with 4 - 5 mm in diameter is fabricated and the maximum traveling velocity of 7 mm/s in both ahead and astern movements is experimentally verified. (author)

Study on flow-induced vibration of large-diameter pipings in a sodium-cooled fast reactor. Influence of elbow curvature on velocity fluctuation field

International Nuclear Information System (INIS)

Ono, Ayako; Kimura, Nobuyuki; Kamide, Hideki; Tobita, Akira

2010-02-01

The main cooling system of Japan Sodium-cooled Fast Reactor (JSFR) consists of two loops to reduce the plant construction cost. In the design of JSFR, sodium coolant velocity is beyond 9m/s in the primary hot leg pipe with large-diameter (1.3m). The maximum Reynolds number in the piping reaches 4.2x10 7 . The hot leg pipe having a 90 degree elbow with curvature ratio of r/D=1.0, so-called 'short elbow', which enables a compact reactor vessel. In sodium cooled fast reactors, the system pressure is so low that thickness of pipings in the cooling system is thinner than that in LWRs. Under such a system condition in the cooling system, the flow-induced vibration (FIV) is concerned at the short elbow. The evaluation of the structural integrity of pipings in JSFR should be conducted based on a mechanistic approach of FIV at the elbow. It is significant to obtain the knowledge of the fluctuation intensity and spectra of velocity and pressure fluctuations in order to grasp the mechanism of the FIV. In this study, water experiments were conducted. Two types of 1/8 scaled elbows with different curvature ratio, r/D=1.0, 1.5, were used to investigate the influence of curvature on velocity fluctuation at the elbow. The velocity fields in the elbows were measured using a high speed PIV method. Unsteady behavior of secondary flow at the elbow outlet and separation flow at the inner wall of elbow were observed in the two types of elbows. It was found that the growth of secondary flow correlated with the flow fluctuation near the inside wall of the elbow. (author)

Probabilistic fracture mechanics analysis for leak-before-break evaluation of light water reactor's piping

International Nuclear Information System (INIS)

Yoshimura, Shinobu; Yagawa, Genki; Akiba, Hiroshi; Fujioka, Terutaka.

1997-01-01

This paper describes Probabilistic Fracture Mechanics (PFM) analyses for quantitative evaluation of the likelihood of Leak-Before-Break (LBB) of Light Water Reactor's (LWR's) piping. The PFM analyses in general assume probabilistic distributions of initial crack size, applied stress cycles, crack growth laws, fracture criteria, leakage detection capability, defect inspection capability and so on. Referring to the deterministic procedure for LBB evaluation, most appropriate PFM models and data for LBB evaluation are discussed. Here the LBB index is newly proposed in order to quantitatively evaluate the likelihood of LBB. Through intensive sensitivity analyses, it is clarified that the LBB is more likely to occur for larger diameter pipe; the performance of leakage detection significantly affects the LBB likelihood; the LBB likelihood increases with plant's aging even conservatively assuming leak detection capability; the R6 method (Category 1, Option 1) for fracture criterion gives very conservative results; and In-Service Inspection (ISI) reduces the increase rate of failure probability than the failure probability itself. (author)

Overview of Prevention for Water Hammer by Check Valve Action in Nuclear Reactor

International Nuclear Information System (INIS)

Kim, Dayong; Yoon, Hyungi; Seo, Kyoungwoo; Kim, Seonhoon

2016-01-01

Water hammer can cause serious damage to pumping system and unexpected system pressure rise in the pipeline. In nuclear reactor, water hammer can influence on the integrity of safety related system. Water hammer in nuclear reactor have been caused by voiding in normally water-filled lines, steam condensation line containing both steam and water, as well as by rapid check valve action. Therefore, this study focuses on the water hammer by check valve among the sources of water hammer occurrence and suggests proper methodology for check valve type selection against water hammer. This study focuses on the water hammer by check valve action among the sources of water hammer occurrence and suggests proper methodology for check valve type selection against water hammer. If the inadvertent pump trip or pipe rupture in high velocity and pressure pipe is predicted, the fast response check valve such as tiled disc, dual disc and nozzle check valve should be installed in the system. If the inadvertent pump trip or pipe rupture in very high velocity and pressure pipe and excessively large revered flow velocity are predicted, the very slowly closing check valve such as controlled closure check valve should be installed in the system

Overview of Prevention for Water Hammer by Check Valve Action in Nuclear Reactor

Energy Technology Data Exchange (ETDEWEB)

Kim, Dayong; Yoon, Hyungi; Seo, Kyoungwoo; Kim, Seonhoon [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2016-10-15

Water hammer can cause serious damage to pumping system and unexpected system pressure rise in the pipeline. In nuclear reactor, water hammer can influence on the integrity of safety related system. Water hammer in nuclear reactor have been caused by voiding in normally water-filled lines, steam condensation line containing both steam and water, as well as by rapid check valve action. Therefore, this study focuses on the water hammer by check valve among the sources of water hammer occurrence and suggests proper methodology for check valve type selection against water hammer. This study focuses on the water hammer by check valve action among the sources of water hammer occurrence and suggests proper methodology for check valve type selection against water hammer. If the inadvertent pump trip or pipe rupture in high velocity and pressure pipe is predicted, the fast response check valve such as tiled disc, dual disc and nozzle check valve should be installed in the system. If the inadvertent pump trip or pipe rupture in very high velocity and pressure pipe and excessively large revered flow velocity are predicted, the very slowly closing check valve such as controlled closure check valve should be installed in the system.

Careful Determination of Inservice Inspection of piping by Computer Analysis in Nuclear Power Plant

International Nuclear Information System (INIS)

Lim, H. T.; Lee, S. L.; Lee, J. P.; Kim, B. C.

1992-01-01

Stress analysis has been performed using computer program ANSYS in the pressurizer surge line in accordance with ASME Sec. III in order to predict possibility of fatigue failure due to thermal stratification phenomena in pipes connected to reactor coolant system of nuclear power plants. Highly vulnerable area to crack generation has been chosen by the analysis of fatigue due to thermal stress in pressurizer surge line. This kind of result can be helpful to choose the location requiring intensive care during inservice inspection of nuclear power plants

Structural integrity of a reinforced concrete structure and a pipe outlet under hydrogen detonation conditions

International Nuclear Information System (INIS)

Saarenheimo, A.; Silde, A.; Calonius, K.

2002-05-01

Structural integrity of a reinforced concrete wall and a pipe penetration under detonation conditions in a selected reactor building room of Olkiluoto BWR were studied. Hydrogen leakage from the pressurised containment to the sur rounding reactor building is possible during a severe accident. Leaked hydrogen tends to accumulate in the reactor building rooms where the leak is located leading to a stable stratification and locally very high hydrogen concentration. If ignited, a possibility to flame acceleration and detonation cannot be ruled out. The structure may survive the peak detonation transient because the eigenperiod of the structure is considerably longer than the duration of the peak detonation. However, the relatively slowly decreasing static type pressure after a peak detonation damages the wall more severely. Elastic deformations in reinforcement are recoverable and cracks in these areas will close after the pressure decrease. But there will be remarkable compression crushing and the static type slowly decreasing over pressure clearly exceeds the loading capacity of the wall. Structural integrity of a pipe outlet was considered also under detonation conditions. The effect of drag forces was taken into account. Damping and strain rate dependence of yield strength were not taken into consideration. The boundary condition at the end of the pipe line model was varied in order to find out the effect of the stiffness of the pipeline outside the calculation model. The calculation model where the lower pipe end is free to move axially, is conservative from the pipe penetration integrity point of view. Even in this conservative study, the highest peak value for the maximum plastic deformation is 3.5%. This is well below the success criteria found in literature. (au)

LWR type reactor

International Nuclear Information System (INIS)

Kato, Kiyoshi.

1993-01-01

A water injection tank in an emergency reactor core cooling system is disposed at a position above a reactor pressure vessel. A liquid phase portion of the water injection tank and an inlet plenum portion in the reactor pressure vessel are connected by a water injection pipe. A gas phase portion of the water injection tank and an upper portion in the reactor pressure vessel are connected by a gas ventilation pipe. Hydraulic operation valves are disposed in the midway of the water injection pipe and the gas ventilation pipe respectively. A pressure conduit is disposed for connecting a discharge port of a main recycling pump and the hydraulic operation valve. In a case where primary coolants are not sent to the main recycling pump by lowering of a liquid level due to loss of coolants or in a case where the main recycling pump is stopped by electric power stoppage or occurrence of troubles, the discharge pressure of the main recycling pump is lowered. Then, the hydraulic operation valve is opened to release the flow channel, then, boric acid water in the water injection tank is sent into the reactor by a falling head, to lead the reactor to a scram state. (I.N.)

Heat-pipe development for the SPAR space-power system

International Nuclear Information System (INIS)

Ranken, W.A.

1981-01-01

The SPAR space power system design is based on a high temperature fast spectrum nuclear reactor that furnishes heat to a thermoelectric conversion system to generate an electrical power output of 100 kW/sub (e)/. An important feature of this design is the use of alkali metal heat pipes to provide redundant, reliable, and low-loss heat transfer at high temperature. Three sets of heat pipes are used in the system. These include sodium/molybdenum heat pipes to transfer heat from the reactor core to the conversion system, potassium/niobium heat pipes to couple the conversion system to the radiator in a redundant manner, and potassium/titanium heat pipes to distribute rejected heat throughout the radiator surface. The designs of these units are discussed and fabrication methods and testing results are described. 12 figures

Seismic analysis of liquid metal reactor piping systems

International Nuclear Information System (INIS)

Wang, C.Y.

1987-01-01

This paper describes the finite-element numerical algorithm and its applications to LMR piping under seismic excitations. A time-history analysis technique using the implicit temporal integration scheme is addressed. A 3-D pipe element is formulated which has eight degrees of freedom per node (three displacements, three rotations, one membrane displacement, and one bending rotation) to account for the hoop, flexural, rotational, and torsional modes of the piping system. Both geometric and material nonlinearities are considered. This algorithm is unconditionally stable and is particularly suited for the seismic analysis. (orig./GL)

Seismic test of high temperature piping for HTGR

International Nuclear Information System (INIS)

Kobatake, Kiyokazu; Midoriyama, Shigeru; Ooka, Yuzi; Suzuki, Michiaki; Katsuki, Taketsugu

1983-01-01

Since the high temperature pipings for the high temperature gas-cooled reactor contain helium gas at 1000 deg C and 40 kgf/cm 2 , the double-walled pipe type consisting of the external pipe serving as the pressure boundary and the internal pipe with heat insulating structure was adopted. Accordingly, their aseismatic design is one of the important subjects. Recently, for the purpose of grasping the vibration characteristics of these high temperature pipings and obtaining the data required for the aseismatic design, two specimens, that is, a double-walled pipe model and a heat-insulating structure, were made, and the vibration test was carried out on them, using a 30 ton vibration table of Kawasaki Heavy Industries Ltd. In the high temperature pipings of the primary cooling system for the multi-purpose, high temperature gas-cooled experimental reactor, the external pipes of 32 B bore as the pressure boundary and the internal pipes of 26 B bore with internal heat insulation consisting of double layers of fiber and laminated metal insulators as the temperature boundary were adopted. The testing method and the results are reported. As the spring constant of spacers is larger and clearance is smaller, the earthquake wave response of double-walled pipes is smaller, and it is more advantageous. The aseismatic property of the heat insulation structure is sufficient. (Kako, I.)

Inspection of secondary cooling system piping of JMTR

International Nuclear Information System (INIS)

Hanawa, Yoshio; Izumo, Hironobu; Fukasaku, Akitomi; Nagao, Yoshiharu; Kawamura, Hiroshi

2008-06-01

Piping condition was inspected form the view point of long term utilization before the renewal work of the secondary cooling system in the JMTR on FY 2008. As the result, it was confirmed that cracks, swellings and exfoliations in inner lining of the piping could be observed, and corrosion, which was reached by piping ingot, or decrease of piping thickness could hardly be observed. It was therefore confirmed that the strength or the functionality of the piping had been maintained by usual operation and maintenance. Repair of inner lining of the piping during the refurbishment of the JMTR is necessary to long term utilization of the secondary cooling system after restart of the JMTR from the view point of preventive maintenance. In addition, a periodic inspection of inner lining condition is necessary after repair of the piping. (author)

Identification of significant problems related to light water reactor piping systems

International Nuclear Information System (INIS)

1980-07-01

Work on the project was divided into three tasks. In Task 1, past surveys of LWR piping system problems and recent Licensee Event Report summaries are studied to identify the significant problems of LWR piping systems and the primary causes of these problems. Pipe cracking is identified as the most recurring problem and is mainly due to the vibration of pipes due to operating pump-pipe resonance, fluid-flow fluctuations, and vibration of pipe supports. Research relevant to the identified piping system problems is evaluated. Task 2 studies identify typical LWR piping systems and the current loads and load combinations used in the design of these systems. Definitions of loads are reviewed. In Task 3, a comparative study is carried out on the use of nonlinear analysis methods in the design of LWR piping systems. The study concludes that the current linear-elastic methods of analysis may not predict accurately the behavior of piping systems under seismic loads and may, under certain circumstances, result in nonconservative designs. Gaps at piping supports are found to have a significant effect on the response of the piping systems

FFTF report: FFTF piping installation and welding techniques

International Nuclear Information System (INIS)

Gilles, J.

1975-01-01

The main sodium piping with a diameter of 16'' or 28 '' is being installed at the FFTF construction site starting in December 1974. The supplier and authority demarcations are: Combustion Engineering supplies the reactor vessel, guard vessel and adjoining pipes and uses the machine welding equipment ''Dimetrics''; for the piping system of the primary and secondary loops the pipes manufactured by Rollmet at HUICO, Pasco, were delivered and prefabricated there, as far as compatible with the installation. ''Astroarc'' welding machines are used by Bechtel for the piping prefabrication in the weld laboratory as well as on site at the construction site. Technical welding problems occurring during the course of the installation at the construction site and several during this time are described. At present 6 weld seams in the reactor and 14 weld seams in the secondary loop are accepted. The requirement exists to carry out as many welds as possible automatically, in order to produce sodium pipe welds of high technical quality and which are reproducible. The welding equipment is described

BWR type reactor system

International Nuclear Information System (INIS)

Morooka, Shin-ichi.

1980-01-01

Purpose: To reduce the internal structure in a reactor by rapidly and efficiently transferring heat generated in a reactor core out of the reactor and eliminating the danger of radiation exposure. Constitution: Steam generated in a pressure vessel is introduced into heat pipe group by inserting the heat pipe group into the steam dome of the pressure vessel. The introduced steam is condensed in the heat pipes to transfer the heat of the steam to the heat pipe group. The transferred heat is transmitted to a heat exchanger provided out of a containment vessel to generate steam to operate a turbine. Thus, it is not necessary to introduce the steam including radioactive substance externally and can remove only the heat so as to carry out the desired purpose. (Kamimura, M.)

Method of measuring reactor water level

International Nuclear Information System (INIS)

Shinohara, Kaoru.

1979-01-01

Purpose: To provide a water level measuring system so that a reactor water level detecting signal can be corrected in correspondence to a recirculation flow, thereby to carry out a correct water level detection in a wide range of the reactor. Method: According to the operation record of a precursor reactor, the ratio Δh of the lowering of the water level due to the recirculation flow is lowered in proportion to the ratiowith respect to the rated differential pressure of the recirculation flow. Accordingly, the flow of recirculation pump is measured by an elbow differential pressure generator utilizing an elbow of a pipe, and the measured value is multiplied by a gain by a ratio setter, and therefter, an addition computation is carried out by an adder for correcting the signal from a water level detector. When the signal from the water level detector is corrected in this manner, the influence of the lowering of the water level due to the recirculation flow can be removed, and an interlocker predetermined in the defined water level can be actuated, thus the influence of the dynamic pressure due to the recirculation flow acting on the instrumental pipe line detecting the reactor water level can be removed effectively. (Yoshino, Y.)

Abrasion behavior of graphite pebble in lifting pipe of pebble-bed HTR

Energy Technology Data Exchange (ETDEWEB)

Shen, Ke; Su, Jiageng [Institute of Nuclear and New Energy Technology, Advanced Nuclear Energy Technology Cooperation Innovation Center, The Key Laboratory of Advanced Nuclear Engineering and Safety, Ministry of Education, Beijing 10084 (China); Zhou, Hongbo [Institute of Nuclear and New Energy Technology, Advanced Nuclear Energy Technology Cooperation Innovation Center, The Key Laboratory of Advanced Nuclear Engineering and Safety, Ministry of Education, Beijing 10084 (China); Chinergy Co., LTD., Beijing 100193 (China); Peng, Wei; Liu, Bing [Institute of Nuclear and New Energy Technology, Advanced Nuclear Energy Technology Cooperation Innovation Center, The Key Laboratory of Advanced Nuclear Engineering and Safety, Ministry of Education, Beijing 10084 (China); Yu, Suyun, E-mail: suyuan@tsinghua.edu.cn [Center for Combustion Energy, The Key Laboratory for Thermal Science and Power Engineering, Ministry of Educations, Tsinghua University, Beijing 10084 (China)

2015-11-15

Highlights: • Quantitative determination of abrasion rate of graphite pebbles in different lifting velocities. • Abrasion behavior of graphite pebble in helium, air and nitrogen. • In helium, intensive collisions caused by oscillatory motion result in more graphite dust production. - Abstract: A pebble-bed high-temperature gas-cooled reactor (pebble-bed HTR) uses a helium coolant, graphite core structure, and spherical fuel elements. The pebble-bed design enables on-line refueling, avoiding refueling shutdowns. During circulation process, the pebbles are lifted pneumatically via a stainless steel lifting pipe and reinserted into the reactor. Inevitably, the movement of the fuel elements as they recirculate in the reactor produces graphite dust. Mechanical wear is the primary source of graphite dust production. Specifically, the sources are mechanisms of pebble–pebble contact, pebble–wall (structural graphite) contact, and fuel handling (pebble–metal abrasion). The key contribution to graphite dust production is from the fuel handling system, particularly from the lifting pipe. During pneumatic lift, graphite pebbles undergo multiple collisions with the stainless steel lifting pipe, thereby causing abrasion of the graphite pebbles and producing graphite dust. The present work explored the abrasion behavior of graphite pebble in the lifting pipe by measuring the abrasion rate at different lifting velocities. The abrasion rate of the graphite pebble in helium was found much higher than those in air and nitrogen. This gas environment effect could be explained by either tribology behavior or dynamic behavior. Friction testing excluded the possibility of tribology reason. The dynamic behavior of the graphite pebble was captured by analysis of the audio waveforms during pneumatic lift. The analysis results revealed unique dynamic behavior of the graphite pebble in helium. Oscillation and consequently intensive collisions occur during pneumatic lift, causing

Shielding considerations for advanced space nuclear reactor systems

International Nuclear Information System (INIS)

Angelo, J.P. Jr.; Buden, D.

1982-01-01

To meet the anticipated future space power needs, the Los Alamos National Laboratory is developing components for a compact, 100 kW/sub e/-class heat pipe nuclear reactor. The reactor uses uranium dioxide (UO 2 ) as its fuel, and is designed to operate around 1500 k. Heat pipes are used to remove thermal energy from the core without the use of pumps or compressors. The reactor heat pipes transfer mal energy to thermoelectric conversion elements that are advanced versions of the converters used on the enormously successful Voyager missions to the outer planets. Advanced versions of this heat pipe reactor could also be used to provide megawatt-level power plants. The paper reviews the status of this advanced heat pipe reactor and explores the radiation environments and shielding requirements for representative manned and unmanned applications

Ring thermal shield piping modification at Pickering Nuclear Generating Station 'A' Unit 1

International Nuclear Information System (INIS)

Brown, R.; Cobanoglu, M.M.

1995-01-01

Each of the four Pickering Nuclear Generating Station A (PNGSA) CANDU units was constructed with its reactor and dump tank surrounded by a concrete Calandria Vault (CV). The Ring Thermal Shield (RTS) system at PNGSA units is a water cooled structure with internal cooling channels with the purpose of attenuating excessive heat flux from the calandria shell to the end shield rings and adjoining concrete (Figure 1). In newer CANDU units the reactor calandria vessel is surrounded by a large water filled shield tank which eliminates the requirement for the RTS system. The RTS structures are situated in the space between the calandria and the vault walls. Each RTS is assembled from eight flat sided carbon steel segments, tilted towards the calandria and supported from the end shield rings. Cooling water to the RTS is supplied by carbon steel cooling pipes with a portion of the pipe run embedded in the vault walls. Flow through each RTS is divided into two independent circuits, having an inlet and an outlet cooling line. There are four locations of RTS inlet and outlet cooling lines. The inlet lines are located at the bottom and the outlet lines at the top of the RTS. The 'L' shaped section of RTS inlet and outlet cooling lines, from the RTS waterbox to the start of embedded portion at the concrete wall, had become defective due to corrosion induced by excessive Moisture levels in the calandria vaults. An on-line leak sealing capability was developed and placed in service in all four PNGSA units. However, a leak found during the 1994 Unit 1 outage was too large,to seal with the current capability, forcing Ontario Hydro (OH) to develop a method to replace the corroded pipes. The repair project was subject to some lofty performance targets. All tools had to be able to withstand dose rates of up to 3000 Rem/hour. These tools, along with procedures and personnel had to successfully repair the RTS system within 6 months otherwise a costly outage extension would result. This

Lightweight Exhaust Manifold and Exhaust Pipe Ducting for Internal Combustion Engines

Science.gov (United States)

Northam, G. Burton (Inventor); Ransone, Philip O. (Inventor); Rivers, H. Kevin (Inventor)

1999-01-01

An improved exhaust system for an internal combustion gasoline-and/or diesel-fueled engine includes an engine exhaust manifold which has been fabricated from carbon- carbon composite materials in operative association with an exhaust pipe ducting which has been fabricated from carbon-carbon composite materials. When compared to conventional steel. cast iron. or ceramic-lined iron paris. the use of carbon-carbon composite exhaust-gas manifolds and exhaust pipe ducting reduces the overall weight of the engine. which allows for improved acceleration and fuel efficiency: permits operation at higher temperatures without a loss of strength: reduces the "through-the wall" heat loss, which increases engine cycle and turbocharger efficiency and ensures faster "light-off" of catalytic converters: and, with an optional thermal reactor, reduces emission of major pollutants, i.e. hydrocarbons and carbon monoxide.

Lining facility for FBR type reactor

International Nuclear Information System (INIS)

Shimano, Kunio.

1991-01-01

In a lining facility for protecting structural material concretes for concrete buildings in an FBR type power plant, sodium-resistant and heat-resistant first and second coating layers are lined at the surface of concretes, and steam releasing materials are disposed between the first and the second coating layers for releasing water contents evaporated from the concretes to the outside. With such a constitution, since there is no structures for welding steel plates to each other as in the prior art, the fabrication is made easy. Further, since cracks of coating materials can be suppressed, reactor safety is improved. (T.M.)

Reactor control device

International Nuclear Information System (INIS)

Kameda, Akiyuki.

1979-01-01

Purpose: To enable three types of controls, that is, level control, scram control and excess reactivity control required for a reactor by a same mechanism by feeding neutron absorber liquid and pressure control gas to several blind pipes provided in the reactor core. Constitution: A plurality of blind pipes are disposed spaced apart in a reactor core and connected by way of injection pipes to a neutron absorber liquid tank. A pressure regulator is connected to the blind pipes, to which pressure control gas is supplied. The neutron absorber liquid used herein consists of sodium, potassium or their alloy, or mercury as a basic substance incorporated with one or more selected from boron, tantalum, rhenium, europium or their compounds. The level control, scram control and excess reactivity control can be attained by moderating the pressure changes in the pressure control gas or by regulating the fluctuation in the liquid level. (Horiughi, T.)

Experimental study on fluid mixing phenomena in T-pipe junction with upstream elbow

International Nuclear Information System (INIS)

Hiroshi Ogawa; Minoru Igarashi; Nobuyuki Kimura; Hideki Kamide

2005-01-01

Full text of publication follows: Temperature fluctuation in fluid causes high cycle thermal fatigue in shroud structure according to its amplitude and frequency. There are still some incidents of thermal fatigue and leakage in light water reactors (Japanese PWR Tomari-2 in 2003, French PWR CIVAUX in 1998), and also in sodium cooled reactors (French FBR Phenix in 1992). Mixing tee is a typical component where temperature fluctuation occurs. Water experiment has been carried out to investigate temperature fluctuation characteristics and flow velocity field in a simple T-pipe junction with straight inlet pipings for main and branch lines; test facility is named as WATLON (Water Experiment on Fluid Mixing in T-pipe with Long Cycle Fluctuation). Here, influence of upstream elbow in the main pipe was studied in the WATLON facility. Elbow can be set near the mixing tee in a real plant. Outlet of the elbow has biased velocity distribution and also the secondary flow, which decays unsteadily. Temperature distribution in the mixing tee was measured by a movable tree with 17 thermocouples and velocity field was measured by Dynamic PIV (high speed particle image velocimetry) with sampling frequency of 200 Hz. Measured temperature showed that fluctuation intensity near the wall was larger in the elbow geometry than in the straight inlet pipes in a case of wall jet (branch flow velocity is smaller than main pipe flow velocity); high intensity region in the elbow case was enlarged around the jet exiting from the branch pipe. The result of flow velocity measurement showed that secondary flow and biased flow velocity distributions due to the elbow influenced bending of the jet exiting from the branch pipe and the temperature fluctuation intensity around the jet. The detailed flow velocity distributions and the secondary flow of upstream elbow can be measured by Dynamic PIV. Influence of such elbow was discussed based on detailed temperature data together with fluctuated velocity

Condensation of steam in horizontal pipes: model development and validation

International Nuclear Information System (INIS)

Szijarto, R.

2015-01-01

This thesis submitted to the Swiss Federal Institute of Technology ETH in Zurich presents the development and validation of a model for the condensation of steam in horizontal pipes. Condensation models were introduced and developed particularly for the application in the emergency cooling system of a Gen-III+ boiling water reactor. Such an emergency cooling system consists of slightly inclined horizontal pipes, which are immersed in a cold water tank. The pipes are connected to the reactor pressure vessel. They are responsible for a fast depressurization of the reactor core in the case of accident. Condensation in horizontal pipes was investigated with both one-dimensional system codes (RELAP5) and three-dimensional computational fluid dynamics software (ANSYS FLUENT). The performance of the RELAP5 code was not sufficient for transient condensation processes. Therefore, a mechanistic model was developed and implemented. Four models were tested on the LAOKOON facility, which analysed direct contact condensation in a horizontal duct

Evaluation of clamp effects on LMFBR piping systems

International Nuclear Information System (INIS)

Jones, G.L.

1980-01-01

Loop-type liquid metal breeder reactor plants utilize thin-wall piping to mitigate through-wall thermal gradients due to rapid thermal transients. These piping loops require a support system to carry the combined weight of the pipe, coolant and insulation and to provide attachments for seismic restraints. The support system examined here utilizes an insulated pipe clamp designed to minimize the stresses induced in the piping. To determine the effect of these clamps on the pipe wall a non-linear, two-dimensional, finite element model of the clamp, insulation and pipe wall was used to determine the clamp/pipe interface load distributions which were then applied to a three-dimensional, finite element model of the pipe. The two-dimensional interaction model was also utilized to estimate the combined clamp/pipe stiffness

Localization of leaks in underground pipes with the application of radioactive isotopes

International Nuclear Information System (INIS)

Klupa, A.; Morawiec, J.

1983-01-01

A method of leaks localization on gas pipe-lines during resistance and tightness tests was elaborated. The leaks were localized using tracer technique. Sodium 24 was used as a tracer for short sections of the pipe-line (up to 30 km). Only 1 m 3 of water with a tracer was introduced into the pipe-line. A measuring probe was also put into pipe-line. All leaks were detected and the method appeared useful. For the longer sections of the pipe-line iodine 131 ought to be used. (A.S.)

A study on the temperature distribution in the hot leg pipe

International Nuclear Information System (INIS)

Choe, Yoon-Jae; Baik, Se-Jin; Jang, Ho-Cheol; Lee, Byung-Jin; Im, In-Young; Ro, Tae-Sun

2003-01-01

In the hot leg pipes of reactor coolant system of the Korean Standard Nuclear Power Plant (KSNP), a non-uniform distribution in temperature has been observed across the cross-section, which is attributed to the non-uniformity of power distribution in the reactor core usually having a peak in the center region, and to the colder coolant bypass flow through the reactor vessel outlet nozzle clearances. As a result, the arithmetic mean temperature of four Resistance Temperature Detectors (RTDs) installed in each hot leg - two in the upper region and two in the lower region around the pipe wall may not correctly represent the actual coolant bulk temperature. It is also believed that there is a skewness in the velocity profile in the hot leg pipe due to the sudden changes in the flow direction and area from the core to the hot leg pipe, through the reactor vessel outlet plenum. These temperature non-uniformity and velocity skewness affect the measurement of the plant parameter such as the reactor coolant flow rate which is calculated by using the bulk temperature of hot leg pipes. A computational analysis has been performed to simulate the temperature and velocity distributions and to evaluate the uncertainty of temperature correction offset in the hot leg pipe. A commercial CFD code, FLUENT, is used for this analysis. The analysis results are compared with the operational data of KSNP and the scaled-down model test data for System 80. From the comparisons, an uncertainty of correction offset is obtained to measure the bulk temperature of hot leg more accurately, which can be also applied to the operating plants, leading to the reduction of temperature measurement uncertainty. Since the uncertainty of temperature in the hot leg pipe is one of major parameters to calculate the uncertainty of the reactor coolant flow rate, the analysis results can contribute to the improvement of the plant performance and safety by reducing the uncertainty of temperature measurement

75 FR 26273 - Certain Seamless Carbon and Alloy Steel Standard, Line, and Pressure Pipe From China

Science.gov (United States)

2010-05-11

...)] Certain Seamless Carbon and Alloy Steel Standard, Line, and Pressure Pipe From China AGENCY: United States... materially injured or threatened with material injury, or the establishment of an industry in the United States is materially retarded, by reason of subsidized and less-than-fair-value imports from China of...

An overview of environmental degradation of materials in nuclear power plant piping systems

International Nuclear Information System (INIS)

Shack, W.J.

1987-08-01

Piping in light water reactor (LWR) power systems is affected by several types of environmental degradation: intergranular stress corrosion cracking (IGSCC) of austenitic stainless steel piping in boiling water reactors (BWRs) has required research, inspection, and mitigation programs that will ultimately cost several billion dollars; erosion-corrosion of carbon steel piping has been observed frequently in the secondary systems of both BWRs and pressurized water reactors (PWRs); the effect of the BWR environment can greatly diminish the design margin inherent in the ASME Section III fatigue design curves for carbon steel piping; and cast stainless steels are subject to embrittlement after extended thermal aging at reactor operating temperatures. These problems are being addressed by wide-ranging research programs in this country and abroad. The purpose of this review is to highlight some of the accomplishments of these programs and to note some of the remaining unanswered questions

The intermittent contact impact problem in piping systems of nuclear reactor

International Nuclear Information System (INIS)

Martin, A.; Ricard, A.; Millard, A.

1981-09-01

The intermittent contact problem is important in many pipe whip studies, specially as to the safety of nuclear reactors. The impact concept adopted is that of instantaneous impact, so that at the time of impact the two impacting structures instantaneously acquire the same velocity in the impact direction. Energy is dissipated by some mechanism whose spatial and temporal scale is small compared to these scales in the discrete model. This dissipation is associated with local plastic deformation. Different solutions are presented for solving this problem. The first one is a generalization of the modal superposition method, when the nonlinearities of the structure are only due to impact between structural components; the other ones are included in a step by step time history and can take in account geometrical non linearities and of behavior. Some industrial applications in nuclear technology are presented

On-line Monitoring of Instrumentation in Research Reactors

International Nuclear Information System (INIS)

2017-12-01

This publication is the result of a benchmarking effort undertaken under the IAEA coordinated research project on improved instrumentation and control (I&C) maintenance techniques for research reactors. It lays the foundation for implementation of on-line monitoring (OLM) techniques and establishment of the validity of those for improved maintenance practices in research reactors for a number of applications such as change to condition based calibration, performance monitoring of process instrumentation systems, detection of process anomalies and to distinguish between process problems/effects and instrumentation/sensor issues. The techniques and guidance embodied in this publication will serve the research reactor community in providing the technical foundation for implementation of OLM techniques. It is intended to be used by Member States to implement I&C maintenance and to improve performance of research reactors.

Reliability of piping system components. Framework for estimating failure parameters from service data

International Nuclear Information System (INIS)

Nyman, R.; Hegedus, D.; Tomic, B.; Lydell, B.

1997-12-01

This report summarizes results and insights from the final phase of a R and D project on piping reliability sponsored by the Swedish Nuclear Power Inspectorate (SKI). The technical scope includes the development of an analysis framework for estimating piping reliability parameters from service data. The R and D has produced a large database on the operating experience with piping systems in commercial nuclear power plants worldwide. It covers the period 1970 to the present. The scope of the work emphasized pipe failures (i.e., flaws/cracks, leaks and ruptures) in light water reactors (LWRs). Pipe failures are rare events. A data reduction format was developed to ensure that homogenous data sets are prepared from scarce service data. This data reduction format distinguishes between reliability attributes and reliability influence factors. The quantitative results of the analysis of service data are in the form of conditional probabilities of pipe rupture given failures (flaws/cracks, leaks or ruptures) and frequencies of pipe failures. Finally, the R and D by SKI produced an analysis framework in support of practical applications of service data in PSA. This, multi-purpose framework, termed 'PFCA'-Pipe Failure Cause and Attribute- defines minimum requirements on piping reliability analysis. The application of service data should reflect the requirements of an application. Together with raw data summaries, this analysis framework enables the development of a prior and a posterior pipe rupture probability distribution. The framework supports LOCA frequency estimation, steam line break frequency estimation, as well as the development of strategies for optimized in-service inspection strategies

Reliability of piping system components. Framework for estimating failure parameters from service data

Energy Technology Data Exchange (ETDEWEB)

Nyman, R [Swedish Nuclear Power Inspectorate, Stockholm (Sweden); Hegedus, D; Tomic, B [ENCONET Consulting GesmbH, Vienna (Austria); Lydell, B [RSA Technologies, Vista, CA (United States)

1997-12-01

This report summarizes results and insights from the final phase of a R and D project on piping reliability sponsored by the Swedish Nuclear Power Inspectorate (SKI). The technical scope includes the development of an analysis framework for estimating piping reliability parameters from service data. The R and D has produced a large database on the operating experience with piping systems in commercial nuclear power plants worldwide. It covers the period 1970 to the present. The scope of the work emphasized pipe failures (i.e., flaws/cracks, leaks and ruptures) in light water reactors (LWRs). Pipe failures are rare events. A data reduction format was developed to ensure that homogenous data sets are prepared from scarce service data. This data reduction format distinguishes between reliability attributes and reliability influence factors. The quantitative results of the analysis of service data are in the form of conditional probabilities of pipe rupture given failures (flaws/cracks, leaks or ruptures) and frequencies of pipe failures. Finally, the R and D by SKI produced an analysis framework in support of practical applications of service data in PSA. This, multi-purpose framework, termed `PFCA`-Pipe Failure Cause and Attribute- defines minimum requirements on piping reliability analysis. The application of service data should reflect the requirements of an application. Together with raw data summaries, this analysis framework enables the development of a prior and a posterior pipe rupture probability distribution. The framework supports LOCA frequency estimation, steam line break frequency estimation, as well as the development of strategies for optimized in-service inspection strategies. 63 refs, 30 tabs, 22 figs.

Two-phase flow structure in large diameter pipes

International Nuclear Information System (INIS)

Smith, T.R.; Schlegel, J.P.; Hibiki, T.; Ishii, M.

2012-01-01

Highlights: ► Local profiles of various quantities measured in large diameter pipe. ► Database for interfacial area in large pipes extended to churn-turbulent flow. ► Flow regime map confirms previous models for flow regime transitions. ► Data will be useful in developing interfacial area transport models for large pipes. - Abstract: Flow in large pipes is important in a wide variety of applications. In the nuclear industry in particular, understanding of flow in large diameter pipes is essential in predicting the behavior of reactor systems. This is especially true of natural circulation Boiling Water Reactor (BWR) designs, where a large-diameter chimney above the core provides the gravity head to drive circulation of the coolant through the reactor. The behavior of such reactors during transients and during normal operation will be predicted using advanced thermal–hydraulics analysis codes utilizing the two-fluid model. Essential to accurate two-fluid model calculations is reliable and accurate computation of the interfacial transfer terms. These interfacial transfer terms can be expressed as the product of one term describing the potential driving the transfer and a second term describing the available surface area for transfer, or interfacial area concentration. Currently, the interfacial area is predicted using flow regime dependent empirical correlations; however the interfacial area concentration is best computed through the use of the one-dimensional interfacial area transport equation (IATE). To facilitate the development of IATE source and sink term models in large-diameter pipes a fundamental understanding of the structure of the two-phase flow is essential. This understanding is improved through measurement of the local void fraction, interfacial area concentration and gas velocity profiles in pipes with diameters of 0.102 m and 0.152 m under a wide variety of flow conditions. Additionally, flow regime identification has been performed to

78 FR 33403 - Transcontinental Gas Pipe Line Company, LLC; Notice of Intent To Prepare an Environmental...

Science.gov (United States)

2013-06-04

... DEPARTMENT OF ENERGY Federal Energy Regulatory Commission [Docket No. PF13-5-000] Transcontinental Gas Pipe Line Company, LLC; Notice of Intent To Prepare an Environmental Assessment for the Planned Leidy Southeast Expansion Project, Request for Comments on Environmental Issues, and Notice of Public Scoping Meetings The staff of the Federal...

77 FR 59391 - Transcontinental Gas Pipe Line Company, LLC; Notice of Intent To Prepare an Environmental...

Science.gov (United States)

2012-09-27

... DEPARTMENT OF ENERGY Federal Energy Regulatory Commission [Docket No. CP12-497-000] Transcontinental Gas Pipe Line Company, LLC; Notice of Intent To Prepare an Environmental Assesment for the Proposed Brandywine Creek Replacement Project; Request for Comments on Environmental Issues; and Notice of Public Scoping Meeting The staff of the Federal...

75 FR 42738 - Transcontinental Gas Pipe Line Company, LLC; Notice of Intent To Prepare an Environmental...

Science.gov (United States)

2010-07-22

... DEPARTMENT OF ENERGY Federal Energy Regulatory Commission [Docket No. PF10-16-000] Transcontinental Gas Pipe Line Company, LLC; Notice of Intent To Prepare an Environmental Assessment for the Planned Mid-Atlantic Connector Expansion Project, Request for Comments on Environmental Issues, and Notice of Public Scoping Meeting July 15, 2010. The...

3D-FE Modeling of 316 SS under Strain-Controlled Fatigue Loading and CFD Simulation of PWR Surge Line

Energy Technology Data Exchange (ETDEWEB)

Mohanty, Subhasish [Argonne National Lab. (ANL), Argonne, IL (United States); Barua, Bipul [Argonne National Lab. (ANL), Argonne, IL (United States); Listwan, Joseph [Argonne National Lab. (ANL), Argonne, IL (United States); Majumdar, Saurin [Argonne National Lab. (ANL), Argonne, IL (United States); Natesan, Ken [Argonne National Lab. (ANL), Argonne, IL (United States)

2017-03-01

In financial year 2017, we are focusing on developing a mechanistic fatigue model of surge line pipes for pressurized water reactors (PWRs). To that end, we plan to perform the following tasks: (1) conduct stress- and strain-controlled fatigue testing of surge-line base metal such as 316 stainless steel (SS) under constant, variable, and random fatigue loading, (2) develop cyclic plasticity material models of 316 SS, (3) develop one-dimensional (1D) analytical or closed-form model to validate the material models and to understand the mechanics associated with 316 SS cyclic hardening and/or softening, (4) develop three-dimensional (3D) finite element (FE) models with implementation of evolutionary cyclic plasticity, and (5) develop computational fluid dynamics (CFD) model for thermal stratification, thermal-mechanical stress, and fatigue of example reactor components, such as a PWR surge line under plant heat-up, cool-down, and normal operation with/without grid-load-following. This semi-annual progress report presents the work completed on the above tasks for a 316 SS laboratory-scale specimen subjected to strain-controlled cyclic loading with constant, variable, and random amplitude. This is the first time that the accurate 3D-FE modeling of the specimen for its entire fatigue life, including the hardening and softening behavior, has been achieved. We anticipate that this work will pave the way for the development of a fully mechanistic-computer model that can be used for fatigue evaluation of safety-critical metallic components, which are traditionally evaluated by heavy reliance on time-consuming and costly test-based approaches. This basic research will not only help the nuclear reactor industry for fatigue evaluation of reactor components in a cost effective and less time-consuming way, but will also help other safety-related industries, such as aerospace, which is heavily dependent on test-based approaches, where a single full-scale fatigue test can cost

77 FR 50465 - Certain Small Diameter Carbon and Alloy Seamless Standard, Line and Pressure Pipe From Romania...

Science.gov (United States)

2012-08-21

...- 795, and the American Petroleum Institute (API) 5L specifications and meeting the physical parameters... 1000 degrees Fahrenheit, at various American Society of Mechanical Engineers (ASME) code stress levels..., line, and pressure pipes and redraw hollows produced, or equivalent, to the American Society for...

76 FR 40717 - Transcontinental Gas Pipe Line Company, LLC; Notice of Intent To Prepare an Environmental...

Science.gov (United States)

2011-07-11

... DEPARTMENT OF ENERGY Federal Energy Regulatory Commission [Docket No. PF11-4-000] Transcontinental Gas Pipe Line Company, LLC; Notice of Intent To Prepare an Environmental Assessment for the Planned Northeast Supply Link Project, Request for Comments on Environmental Issues, and Notice of Public Scoping Meetings The staff of the Federal Energy...

Characterisation of girth pipe weld for primary heat transport system of pressurised heavy water reactors

International Nuclear Information System (INIS)

Singh, P.K.; Vaze, K.K.; Kushwaha, H.S.

2002-01-01

The weld and heat affected zone (HAZ) associated with the girth weld are most vulnerable regions of the piping system. The different regions of the weld joint such as the weld metal, HAZ and base metal lead to heterogeneous mechanical and metallurgical properties of the joints. Due to their different metallurgical and mechanical properties, the amounts of damage produced in these regions are different when the component is subjected to service condition. Thus, it is imperative to know the characteristics of these regions of a pipe weld in order to identify the weakest zone for safe designing of high energy piping components. In view of this necessity the present study has been planned to carry out complete characterisation of the weld joint of SA 333 Gr.6 steel pipe, in terms of its metallurgical, mechanical and fracture properties. The mechanical and fracture mechanics properties of the base metal, weld deposit and HAZ have been compared and correlated with reference to their microstructures. Weld joints of SA 333 Gr.6 steel pipe have been prepared by using GTAW root pass and SMAW filling of V-grove as per recommended welding procedure specifications (WPS) conforming to ASME Sec IX commonly used to fabricate nuclear piping system components. The emphasis of the study is to characterise base, weld and HAZ of the pipe weld in terms of chemical, metallurgical, mechanical and fracture mechanics properties. The fracture toughness behaviour of the welds and HAZ has been characterised by J-integral parameters. The fatigue crack growth rate has been characterised by Paris Law. Stretched zone width (SZW) has been measured under SEM to evaluate initiation fracture toughness. The estimated initiation fracture toughness based on SZW and blunting line given by EGF recommendation have been compared. The fracture mechanics properties of base, weld and HAZ has been determined and compared. The fracture mechanics properties of the weld and HAZ have been correlated to their

Failure pressure of straight pipe with wall thinning under internal pressure

International Nuclear Information System (INIS)

Kamaya, Masayuki; Suzuki, Tomohisa; Meshii, Toshiyuki

2008-01-01

The failure pressure of pipe with wall thinning was investigated by using three-dimensional elastic-plastic finite element analyses (FEA). With careful modeling of the pipe and flaw geometry in addition to a proper stress-strain relation of the material, FEA could estimate the precise burst pressure obtained by the tests. FEA was conducted by assuming three kinds of materials: line pipe steel, carbon steel, and stainless steel. The failure pressure obtained using line pipe steel was the lowest under the same flaw size condition, when the failure pressure was normalized by the value of unflawed pipe defined using the flow stress. On the other hand, when the failure pressure was normalized by the results of FEA obtained for unflawed pipe under various flaw and pipe configurations, the failure pressures of carbon steel and line pipe steel were almost the same and lower than that of stainless steel. This suggests that the existing assessment criteria developed for line pipe steel can be applied to make a conservative assessment of carbon steel and stainless steel

Design considerations for CRBRP heat transport system piping operating at elevated temperatures

International Nuclear Information System (INIS)

Pollono, L.P.; Mello, R.M.

1979-01-01

The heat transport system sodium piping for the Clinch River Breeder Reactor Plant (CRBRP) within the reactor containment building must withstand high temperatures for long periods of time. Each phase of the mechanical design process of the piping system is influenced by elevated temperature considerations which include material thermal creep effects, ratchetting caused by rapid temperature transients and stress relaxation, and material degradation effects. The structural design philosophy taken to design the CRBRP piping operating in a high temperature environment is described. The resulting design of the heat transport system piping is presented along with a discussion of special features that resulted from the elevated temperature considerations

Probabilistic pipe fracture evaluations for leak-rate-detection applications

International Nuclear Information System (INIS)

Rahman, S.; Ghadiali, N.; Paul, D.; Wilkowski, G.

1995-04-01

Regulatory Guide 1.45, open-quotes Reactor Coolant Pressure Boundary Leakage Detection Systems,close quotes was published by the U.S. Nuclear Regulatory Commission (NRC) in May 1973, and provides guidance on leak detection methods and system requirements for Light Water Reactors. Additionally, leak detection limits are specified in plant Technical Specifications and are different for Boiling Water Reactors (BWRs) and Pressurized Water Reactors (PWRs). These leak detection limits are also used in leak-before-break evaluations performed in accordance with Draft Standard Review Plan, Section 3.6.3, open-quotes Leak Before Break Evaluation Proceduresclose quotes where a margin of 10 on the leak detection limit is used in determining the crack size considered in subsequent fracture analyses. This study was requested by the NRC to: (1) evaluate the conditional failure probability for BWR and PWR piping for pipes that were leaking at the allowable leak detection limit, and (2) evaluate the margin of 10 to determine if it was unnecessarily large. A probabilistic approach was undertaken to conduct fracture evaluations of circumferentially cracked pipes for leak-rate-detection applications. Sixteen nuclear piping systems in BWR and PWR plants were analyzed to evaluate conditional failure probability and effects of crack-morphology variability on the current margins used in leak rate detection for leak-before-break

Reliability of piping system components. Volume 4: The pipe failure event database

Energy Technology Data Exchange (ETDEWEB)

Nyman, R; Erixon, S [Swedish Nuclear Power Inspectorate, Stockholm (Sweden); Tomic, B [ENCONET Consulting GmbH, Vienna (Austria); Lydell, B [RSA Technologies, Visat, CA (United States)

1996-07-01

Available public and proprietary databases on piping system failures were searched for relevant information. Using a relational database to identify groupings of piping failure modes and failure mechanisms, together with insights from published PSAs, the project team determined why, how and where piping systems fail. This report represents a compendium of technical issues important to the analysis of pipe failure events, and statistical estimation of failure rates. Inadequacies of traditional PSA methodology are addressed, with directions for PSA methodology enhancements. A `data driven and systems oriented` analysis approach is proposed to enable assignment of unique identities to risk-significant piping system component failure. Sufficient operating experience does exist to generate quality data on piping failures. Passive component failures should be addressed by today`s PSAs to allow for aging analysis and effective, on-line risk management. 42 refs, 25 figs.

Reliability of piping system components. Volume 4: The pipe failure event database

International Nuclear Information System (INIS)

Nyman, R.; Erixon, S.; Tomic, B.; Lydell, B.

1996-07-01

Available public and proprietary databases on piping system failures were searched for relevant information. Using a relational database to identify groupings of piping failure modes and failure mechanisms, together with insights from published PSAs, the project team determined why, how and where piping systems fail. This report represents a compendium of technical issues important to the analysis of pipe failure events, and statistical estimation of failure rates. Inadequacies of traditional PSA methodology are addressed, with directions for PSA methodology enhancements. A 'data driven and systems oriented' analysis approach is proposed to enable assignment of unique identities to risk-significant piping system component failure. Sufficient operating experience does exist to generate quality data on piping failures. Passive component failures should be addressed by today's PSAs to allow for aging analysis and effective, on-line risk management. 42 refs, 25 figs

Development of high-strength heavy-wall sour-service seamless line pipe for deep water by applying inline heat treatment

Energy Technology Data Exchange (ETDEWEB)

Arai, Y.; Kondo, K.; Hamada, M.; Hisamune, N.; Murao, N.; Murase, T.; Osako, H. [Sumitomo Metal Industries Ltd., Tokyo (Japan)

2004-07-01

This paper provided details of a new high-strength heavy-wall sour service seamless line pipe developed for use in deep water applications. Pig iron was processed in a blast furnace and refined. Molten steel was degassed to reduce impurities and poured into a continuous caster with a round mold. Billets were then heated in a walking-beam furnace and then pierced to form a hollow shell. The shell was then rolled to a specific thickness in a compact mandrel mill and rolled to a specified outer diameter by an extracting sizer. A heating furnace was used to improve the uniformity of the pipes. The heated pipes were then moved to a cooling zone, then rotated quickly while a high-pressured jet flow was injected inside the pipe at the same time as a slit laminar flow was applied to the outside of the pipe. Higher strength was achieved by using the high performance quenching device. It was noted that while pipes manufactured using the inline heat treatment process were able to achieve higher strengths, toughness was reduced. Metallurgical tests were conducted to improve the toughness value of the seamless pipe. Both the microstructure and the fracture surface of test specimens were examined using scanning electron microscopy. Results of the tests showed that lowering sulphur (S) and titanium (Ti) content improved the toughness properties of the pipes. It was concluded that control of microalloys is important to secure improved toughness for pipes manufactured using inline heat treatments. 5 tabs., 12 figs.

Piping engineering and operation

International Nuclear Information System (INIS)

1993-01-01

The conference 'Piping Engineering and Operation' was organized by the Institution of Mechanical Engineers in November/December 1993 to follow on from similar successful events of 1985 and 1989, which were attended by representatives from all sectors of the piping industry. Development of engineering and operation of piping systems in all aspects, including non-metallic materials, are highlighted. The range of issues covered represents a balance between current practices and implementation of future international standards. Twenty papers are printed. Two, which are concerned with pressurized pipes or steam lines in the nuclear industry, are indexed separately. (Author)

Studies of S-CO{sub 2} Power Plant Pipe Design for Small Modular Sodium-cooled Fast Reactor

Energy Technology Data Exchange (ETDEWEB)

Kim, Min Seok; Ahn, Yoon Han; Lee, Jeong Ik [Korea Advanced Institute of Science and Technology, Daejeon (Korea, Republic of)

2014-10-15

If SFR can be developed into the economical small modular reactor (SMR) for an export from Korea, the expected value can be greater. However, current SFR design may face difficulty in public acceptance due to the potential hazard from sodium-water reaction (SWR) when the current conventional steam Rankine cycle is utilized as a power conversion system for a SFR. In order to eliminate SWR, the Supercritical CO{sub 2} (S-CO{sub 2}) cycle has been proposed. Although there are many researches on S-CO{sub 2} cycle concept and turbomachinery, very few research works considered pipe selection criteria for the S-CO{sub 2} cycle. As one of the most important parts of the plant, this paper will discuss how to select a suitable pipe considering thermal expansion for the S-CO{sub 2} power plant and perform a conceptual design of SFR type SMR. The S-CO{sub 2} cycle can improve the safety of SFR as preventing the SWR by changing the working fluid. Additionally, not only the relatively high efficiency with 450-750 .deg. C turbine inlet temperature, but also the physically compact footprint are advantages of the S-CO{sub 2} cycle. However the pipe design is more complicated than existing power plant because it has high pressure and temperature conditions and needs high mass flow rate. By designing the piping system for a small modular -SFR, the compactness and simplicity of the S-CO{sub 2} cycle are re-confirmed. Moreover, in this paper, realistic and safe pipe design was conducted by considering thermal expansion in the high pressure and temperature conditions. Although total pipe pressure drop is somewhat high, the cycle thermal efficiency is still higher than the existing steam Rankine cycle. Additional study for a larger system such as 300MW class system in MIT report will be conducted in the future study. From the preliminary estimation when the S-CO{sub 2} system becomes large, the pipe diameter may exceed the current ASME standard. This means that more innovative approach

LBB application in the US operating and advanced reactors

Energy Technology Data Exchange (ETDEWEB)

Wichman, K.; Tsao, J.; Mayfield, M.

1997-04-01

The regulatory application of leak before break (LBB) for operating and advanced reactors in the U.S. is described. The U.S. Nuclear Regulatory Commission (NRC) has approved the application of LBB for six piping systems in operating reactors: reactor coolant system primary loop piping, pressurizer surge, safety injection accumulator, residual heat removal, safety injection, and reactor coolant loop bypass. The LBB concept has also been applied in the design of advanced light water reactors. LBB applications, and regulatory considerations, for pressurized water reactors and advanced light water reactors are summarized in this paper. Technology development for LBB performed by the NRC and the International Piping Integrity Research Group is also briefly summarized.

Significance assessment of small-medium sized reactors

Energy Technology Data Exchange (ETDEWEB)

Kanno, Minoru [Japan Atomic Power Co., Research and Development Dept., Tokyo (Japan)

2002-12-01

Preliminary assessment for deployment of small-medium sized reactor (S and M reactor) as a future option has been conducted at the JAPCO (Japan Atomic Power Company) under the cooperation with the CRIERI (Central Research Institute of Electric Power Industry). Significance of the S and M reactor introduction is listed as follows; lower investment cost, possible siting near demand side, enlarged freedom of siting, shorter transmission line, good compatibility with slow increase of demand and plain explanation of safety using simpler system such as integral type vessel without piping, natural convection core cooling and passive safety system. The deployment of simpler plant system, modular shop fabrication, ship-shell structured building and longer operation period can assure economics comparable with that of a large sized reactor, coping with scale-demerit. Also the S and M reactor is preferable in size for the nuclear heat utilization such as hydrogen production. (T. Tanaka)

Pipe fracture evaluations for leak-rate detection: Probabilistic models

International Nuclear Information System (INIS)

Rahman, S.; Wilkowski, G.; Ghadiali, N.

1993-01-01

This is the second in series of three papers generated from studies on nuclear pipe fracture evaluations for leak-rate detection. This paper focuses on the development of novel probabilistic models for stochastic performance evaluation of degraded nuclear piping systems. It was accomplished here in three distinct stages. First, a statistical analysis was conducted to characterize various input variables for thermo-hydraulic analysis and elastic-plastic fracture mechanics, such as material properties of pipe, crack morphology variables, and location of cracks found in nuclear piping. Second, a new stochastic model was developed to evaluate performance of degraded piping systems. It is based on accurate deterministic models for thermo-hydraulic and fracture mechanics analyses described in the first paper, statistical characterization of various input variables, and state-of-the-art methods of modem structural reliability theory. From this model. the conditional probability of failure as a function of leak-rate detection capability of the piping systems can be predicted. Third, a numerical example was presented to illustrate the proposed model for piping reliability analyses. Results clearly showed that the model provides satisfactory estimates of conditional failure probability with much less computational effort when compared with those obtained from Monte Carlo simulation. The probabilistic model developed in this paper will be applied to various piping in boiling water reactor and pressurized water reactor plants for leak-rate detection applications

Pressure tube reactor

International Nuclear Information System (INIS)

Seki, Osamu; Kumasaka, Katsuyuki.

1988-01-01

Purpose: To remove the heat of reactor core using a great amount of moderators at the periphery of the reactor core as coolants. Constitution: Heat of a reactor core is removed by disposing a spontaneous recycling cooling device for cooling moderators in a moderator tank, without using additional power driven equipments. That is, a spontaneous recycling cooling device for cooling the moderators in the moderator tank is disposed. Further, the gap between the inner wall of a pressure tube guide pipe disposed through the vertical direction of a moderator tank and the outer wall of a pressure tube inserted through the guide pipe is made smaller than the rupture distortion caused by the thermal expansion upon overheating of the pressure tube and greater than the minimum gap required for heat shiels between the pressure tube and the pressure tube guide pipe during usual operation. In this way, even if such an accident as can not using a coolant cooling device comprising power driven equipment should occur in the pressure tube type reactor, the rise in the temperature of the reactor core can be retarded to obtain a margin with time. (Kamimura, M.)

Radioactivity Monitoring System for TRIGA 2000 Reactor Water Tank with On-Line Gamma Spectrometer

International Nuclear Information System (INIS)

Prasetyo Basuki; Sudjatmi KA

2009-01-01

One of the requirements in radiological safety in the operating condition of research reactor are the absence of radionuclide from fission product released to reactor cooling water and environment. Early detection of fission product that released from fuel element can be done by monitoring radioactivity level on primary cooling water.Reactor cooling water can be used as an important indicator in detecting radioactivity level of material fission product, when the leakage occurs. Therefore, it needs to build a monitoring system for measuring radioactivity level of cooling water directly and simple. The idea of this system is counting radioactivity water flow from reactor tank to the marinelli cube that attached to the HPGe detector on gamma spectrometer. Cooling water from tank aimed on plastic pipe to the marinelli cube. Water flows in gravitational driven to the marinelli cube, with volume flow rate 5.1 liters/minute in the inlet and 2.2 liters/minute in output. (author)

Effect of pipe rupture loads inside containment in the break exclusionary piping outside containment

International Nuclear Information System (INIS)

Weiss, G.

1987-01-01

The plant design for protection against piping failures outside containment should make sure that fluid system piping in containment penetration areas are designed to meet the break exclusionary provisions contained in the BTP MEB 3-1. According to these provisions, following a piping failure (main steam line) inside containment, the part of the flued head connected to the piping outside containment, should not exceed the ASME Code stress limits for the appropriate load combinations. A finite element analysis has been performed to evaluate the stress level in this area. (orig./HP)

Support structure for reactor core constituent element

International Nuclear Information System (INIS)

Aida, Yasuhiko.

1993-01-01

A connection pipe having an entrance nozzle inserted therein as a reactor core constituent element is protruded above the upper surface of a reactor core support plate. A through hole is disposed to the protruding portion of the connection pipe. The through hole and a through hole disposed to the reactor core support plate are connected by a communication pipe. A shear rod is disposed in a horizontal portion at the inside of the communication pipe and is supported by a spring horizontally movably. Further, a groove is disposed at a position of the entrance nozzle opposing to the shear rod. The shear rod is urged out of the communication pipe by the pressure of the high pressure plenum and the top end portion of the shear rod is inserted to the groove of the entrance nozzle during operation. Accordingly, the shear rod is positioned in a state where it is extended from the through hole of the communication pipe to the groove of the entrance nozzle. This can mechanically constrain the rising of the reactor core constituent elements by the shear rod upon occurrence of earthquakes. (I.N.)

Inspection technology for high pressure pipes

Energy Technology Data Exchange (ETDEWEB)

Kim, Jae H.; Lee, Jae C.; Eum, Heung S.; Choi, Yu R.; Moon, Soon S.; Jang, Jong H

2000-02-01

Various kinds of defects are likely to be occurred in the welds of high pressure pipes in nuclear power plants. Considering the recent accident of Zuruga nuclear power plant in Japan, reasonable policy is strongly requested for the high pressure pipe integrity. In this study, we developed the technologies to inspect pipe welds automatically. After development of scanning robot prototype in the first research year, we developed and implemented the algorithm of automatic tracking of the scanning robot along the weld line of the pipes. We use laser slit beam on weld area and capture the image using digital camera. Through processing of the captures image, we finally determine the weld line automatically. In addition, we investigated a new technology on micro systems for developing micro scanning robotic inspection of the pipe welds. The technology developed in this study is being transferred to the industry. (author)

B Plant process piping replacement feasibility study

International Nuclear Information System (INIS)

Howden, G.F.

1996-01-01

Reports on the feasibility of replacing existing embedded process piping with new more corrosion resistant piping between cells and between cells and a hot pipe trench of a Hanford Site style canyon facility. Provides concepts for replacement piping installation, and use of robotics to replace the use of the canyon crane as the primary means of performing/supporting facility modifications (eg, cell lining, pipe replacement, equipment reinstallation) and operational maintenenace

Application of tearing modulus stability concepts to nuclear piping. Final report

International Nuclear Information System (INIS)

Cotter, K.H.; Chang, H.Y.; Zahoor, A.

1982-02-01

The recently developed tearing modulus stability concept was successfully applied to several boiling water reactor (BWR) and pressurized water reactor (PWR) piping systems. Circumferentially oriented through-the-thickness cracks were postulated at numerous locations in each system. For each location, the simplified tearing stability methods developed in USNRC Report NUREG/CR-0838 were used to determine crack stability. The J-T diagram was used to present the results of the computations. The piping systems considered included Type 304 stainless steel as well as A106 carbon steel materials. These systems were analyzed using the piping analysis computer code MINK

Development of bore tools for pipe welding and cutting

International Nuclear Information System (INIS)

Oka, Kiyoshi; Ito, Akira; Takiguchi, Yuji

1998-01-01

In the International Thermonuclear Experimental Reactor (ITER), in-vessel components replacement and maintenance requires that connected cooling pipes be cut and removed beforehand and that new components be installed to which cooling pipes must be rewelded. All welding must be inspected for soundness after completion. These tasks require a new task concept for ensuring shielded areas and access from narrow ports. Thus, it became necessary to develop autonomous locomotion welding and cutting tools for branch and main pipes to weld pipes by in-pipe access; a system was proposed that cut and welded branch and main pipes after passing inside pipe curves, and elemental technologies developed. This paper introduces current development in tools for welding and cutting branch pipes and other tools for welding and cutting the main pipe. (author)

Development of bore tools for pipe welding and cutting

Energy Technology Data Exchange (ETDEWEB)

Oka, Kiyoshi; Ito, Akira; Takiguchi, Yuji [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

1998-04-01

In the International Thermonuclear Experimental Reactor (ITER), in-vessel components replacement and maintenance requires that connected cooling pipes be cut and removed beforehand and that new components be installed to which cooling pipes must be rewelded. All welding must be inspected for soundness after completion. These tasks require a new task concept for ensuring shielded areas and access from narrow ports. Thus, it became necessary to develop autonomous locomotion welding and cutting tools for branch and main pipes to weld pipes by in-pipe access; a system was proposed that cut and welded branch and main pipes after passing inside pipe curves, and elemental technologies developed. This paper introduces current development in tools for welding and cutting branch pipes and other tools for welding and cutting the main pipe. (author)

Resolution of thermal striping issue downstream of a horizontal pipe elbow in stratified pipe flow

International Nuclear Information System (INIS)

Kuzay, T.M.; Kasza, K.E.

1985-01-01

A thermally stratified pipe flow produced by a thermal transient when passing through a horizontal elbow as a result of secondary flow gives rise to large thermal fluctuations on the inner curvature wall of the downstream piping. These fluctuations were measured in a specially instrumented horizontal pipe and elbow system on a test set-up using water in the Mixing Components Technology Facility (MCTF) at Argonne National Laboratory (ANL). This study is part of a larger program which is studying the influence of thermal buoyancy on general reactor component performance. This paper discusses the influence of pipe flow generated thermal oscillations on the thermal stresses induced in the pipe walls. The instrumentation was concentrated around the exit plane of the 90 0 sweep elbow, since prior tests had indicated that the largest thermal fluctuations would occur within about one hydraulic diameter downstream of the elbow exit. The thermocouples were located along the inner curvature of the piping and measured the near surface fluid temperature. The test matrix involved thermal downramps under turbulent flow conditions

77 FR 56809 - Certain Small Diameter Seamless Carbon and Alloy Steel Standard, Line, and Pressure Pipe From...

Science.gov (United States)

2012-09-14

... stress levels. Alloy pipes made to ASTM standard A-335 must be used if temperatures and stress levels... not used in standard, line or pressure applications. In addition, finished and unfinished oil country... antidumping duty order from the same country. If not covered by such an OCTG order, finished and unfinished...

75 FR 69050 - Certain Seamless Carbon and Alloy Steel Standard, Line, and Pressure Pipe From the People's...

Science.gov (United States)

2010-11-10

... amended (``the Act''), that an industry in the United States is threatened with material injury by reason... case is based on the threat of material injury and is not accompanied by a finding that injury would... Alloy Steel Standard, Line, and Pressure Pipe From the People's Republic of China: Amended Final...

Reactor technology. Progress report, January--March 1978

International Nuclear Information System (INIS)

Warren, J.L.

1978-07-01

Progress is reported in eight program areas. The nuclear Space Electric Power Supply Program examined safety questions in the aftermath of the COSMOS 954 incident, examined the use of thermoelectric converters, examined the neutronic effectiveness of various reflecting materials, examined ways of connecting heat pipes to one another, studied the consequences of the failure of one heat pipe in the reactor core, and did conceptual design work on heat radiators for various power supplies. The Heat Pipe Program reported progress in the design of ceramic heat pipes, new application of heat pipes to solar collectors, and final performance tests of two pipes for HEDL applications. Under the Nuclear Process Heat Program, work continues on computer codes to model a pebble bed high-temperature gas-cooled reactor, adaptation of a set of German reactor calculation codes to use on U.S. computers, and a parametric study of a certain resonance integral required in reactor studies. Under the Nonproliferation Alternative Sources Assessment Program LASL has undertaken an evaluation of a study of gaseous core reactors by Southern Science Applications, Inc. Independently LASL has developed a proposal for a comprehensive study of gaseous uranium-fueled reactor technology. The Plasma Core Reactor Program has concentrated on restacking the beryllium reflector and redesigning the nuclear control system. The status of and experiments on four critical assemblies, SKUA, Godiva IV, Big Ten, and Flattop, are reported. The Nuclear Criticality Safety Program carried out several tasks including conducting a course, doing several annual safety reviews and evaluating the safety of two Nevada test devices. During the quarter one of the groups involved in reactor technology has acquired responsibility for the operation of a Cockroft-Walton accelerator. The present report contains information on the use of machine and improvements being made in its operation

Development of bore tools for pipe inspection

Energy Technology Data Exchange (ETDEWEB)

Oka, Kiyoshi; Nakahira, Masataka; Taguchi, Kou; Ito, Akira [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

1998-04-01

In the International Thermonuclear Reactor (ITER), replacement and maintenance on in-vessel components requires that all cooling pipes connected be cut and removed, that a new component be installed, and that all cooling pipes be rewelded. After welding is completed, welded area must be inspected for soundness. These tasks require a new work concept for securing shielded area and access from narrow ports. Tools had to be developed for nondestructive inspection and leak testing to evaluate pipe welding soundness by accessing areas from inside pipes using autonomous locomotion welding and cutting tools. A system was proposed for nondestructive inspection of branch pipes and the main pipe after passing through pipe curves, the same as for welding and cutting tool development. Nondestructive inspection and leak testing sensors were developed and the basic parameters were obtained. In addition, the inspection systems which can move inside pipes and conduct the nondestructive inspection and the leak testing were developed. In this paper, an introduction will be given to the current situation concerning the development of nondestructive inspection and leak testing machines for the branch pipes. (author)

Recycling systems for BWR type reactors

International Nuclear Information System (INIS)

Takagi, Akio; Yamamoto, Fumiaki; Fukumoto, Ryuji.

1986-01-01

Purpose: To stabilize the coolant flowing characteristics and reactor core reactivity. Constitution: The recycling system in a BWR type reactor comprises a recycling pump disposed to the outside of a reactor pressure vessel, a ring header connected to the recycling pump through main pipe ways, and a plurality of pipes branched from and connected with the ring header and connected to a plurality of jet pumps within the pressure vessel. Then, by making the diameter for the pipeways of each of the branched pipes different from each other, the effective cross-sectional area is varied to thereby average the coolant flow rate supplied to each of the jet pumps. (Seki, T.)

Numerical simulations of pressure fluctuations at branch piping in BWR main steam line

International Nuclear Information System (INIS)

Morita, Ryo; Inada, Fumio; Yoshikawa, Kazuhiro; Takahashi, Shiro

2009-01-01

The power uprating of a nuclear power plant may increase/accelerate degradation phenomena such as flow-induced vibration and wall thinking. A steam dryer was damaged by a high cycle fatigue due to an acoustic-induced vibration at the branch piping of safety relief valves (SRVs) in main steam lines. In this study, we conducted the numerical simulations of steam/air flow around a simplified branch piping to clarify the basic characteristics of resonance. LES simulations were conducted in ordinary pressure/temperature air and steam under BWR plant conditions. In both cases, the excitation of the pressure fluctuations at the branch was observed under some inlet velocity conditions. These fluctuations and inlet conditions were normalized and the obtained results were compared. The normalized results showed that the range and maximum amplitude of pressure fluctuations were almost the same in low-pressure/temperature air and high-pressure/temperature steam. We found that ordinary pressure/temperature air experiments and simulations can possibly clarify the characteristics of the resonance in high-pressure/temperature steam. (author)

77 FR 32626 - Transcontinental Gas Pipe Line Company, LLC; Notice of Intent To Prepare an Environmental Impact...

Science.gov (United States)

2012-06-01

... DEPARTMENT OF ENERGY Federal Energy Regulatory Commission [Docket No. PF09-8-000] Transcontinental Gas Pipe Line Company, LLC; Notice of Intent To Prepare an Environmental Impact Statement for the Planned Rockaway Delivery Lateral Project, Request for Comments on Environmental Issues, and Notice of Public Scoping Meetings The staff of the Federa...

Intergranular stress corrosion cracking: A rationalization of apparent differences among stress corrosion cracking tendencies for sensitized regions in the process water piping and in the tanks of SRS reactors

International Nuclear Information System (INIS)

Louthan, M.R.

1990-01-01

The frequency of stress corrosion cracking in the near weld regions of the SRS reactor tank walls is apparently lower than the cracking frequency near the pipe-to-pipe welds in the primary cooling water system. The difference in cracking tendency can be attributed to differences in the welding processes, fabrication schedules, near weld residual stresses, exposure conditions and other system variables. This memorandum discusses the technical issues that may account the differences in cracking tendencies based on a review of the fabrication and operating histories of the reactor systems and the accepted understanding of factors that control stress corrosion cracking in austenitic stainless steels

An overview of environmental degradation of materials in nuclear power plant piping systems

International Nuclear Information System (INIS)

Shack, W.J.

1988-01-01

Several types of environmental degradation of piping in light water reactor (LWR) power systems have already had significant economic impact on the industry. These include intergranular stress corrosion cracking (IGSCC) of austenitic stainless steel piping, erosion-corrosion of carbon steel piping in secondary systems, and a variety of types of fatigue failures. In addition, other problems have been identified that must be addressed in considering extended lifetimes for nuclear plants. These include the embrittlement of cast stainless steels after extended thermal aging at reactor operating temperatures and the effect of reactor environments on the design margin inherent in the ASME Section III fatigue design curves especially for carbon steel piping. These problems are being addressed by wide-ranging research programs in this country and abroad. The purpose of this review is to highlight some of the accomplishments of these programs and to note some of the remaining unanswered questions

Noncondensable gas accumulation phenomena in nuclear power plant piping

International Nuclear Information System (INIS)

Yamamoto, Yasushi; Aoki, Kazuyoshi; Sato, Teruaki; Shida, Akira; Ichikawa, Nagayoshi; Nishikawa, Akira; Inagaki, Tetsuhiko

2011-01-01

In the case of the boiling water reactor, hydrogen and oxygen slightly exist in the main steam, because these noncondensable gases are generated by the radiolytic decomposition of the reactor water. BWR plants have taken measures to prevent noncondensable gas accumulation. However, in 2001, the detonation of noncondensable gases occurred at Hamaoka-1 and Brunsbuttel, resulting in ruptured piping. The accumulation phenomena of noncondensable gases in BWR closed piping must be investigated and understood in order to prevent similar events from occurring in the future. Therefore, an experimental study on noncondensable gas accumulation was carried out. The piping geometries for testing were classified and modeled after the piping of actual BWR plants. The test results showed that 1) noncondensable gases accumulate in vertical piping, 2) it is hard for noncondensable gases to accumulate in horizontal piping, and 3) noncondensable gases accumulate under low-pressure conditions. A simple accumulation analysis method was proposed. To evaluate noncondensable gas accumulation phenomena, the three component gases were treated as a mixture. It was assumed that the condensation amount of the vapor is small, because the piping is certainly wrapped with heat insulation material. Moreover, local thermal equilibrium was assumed. This analysis method was verified using the noncondensable gas accumulation test data on branch piping with a closed top. Moreover, an experimental study on drain trap piping was carried out. The test results showed that the noncondensable gases dissolved in the drain water were discharged from the drain trap, and Henry's law could be applied to evaluate the amount of dissolved noncondensable gases in the drain water. (author)

Modeling of pipe break accident in a district heating system using RELAP5 computer code

International Nuclear Information System (INIS)

Kaliatka, A.; Valinčius, M.

2012-01-01

Reliability of a district heat supply system is a very important factor. However, accidents are inevitable and they occur due to various reasons, therefore it is necessary to have possibility to evaluate the consequences of possible accidents. This paper demonstrated the capabilities of developed district heating network model (for RELAP5 code) to analyze dynamic processes taking place in the network. A pipe break in a water supply line accident scenario in Kaunas city (Lithuania) heating network is presented in this paper. The results of this case study were used to demonstrate a possibility of the break location identification by pressure decrease propagation in the network. -- Highlights: ► Nuclear reactor accident analysis code RELAP5 was applied for accident analysis in a district heating network. ► Pipe break accident scenario in Kaunas city (Lithuania) district heating network has been analyzed. ► An innovative method of pipe break location identification by pressure-time data is proposed.

Quality control of stainless steel pipings for nuclear power generation

International Nuclear Information System (INIS)

Miki, Minoru; Kitamura, Ichiro; Ito, Hisao; Sasaki, Ryoichi

1979-01-01

The proportion of nuclear power in total power generation is increasing recently in order to avoid the concentrated dependence on petroleum resources, consequently the reliability of operation of nuclear power plants has become important. In order to improve the reliability of plants, the reliability of each machine or equipment must be improved, and for the purpose, the quality control at the time of manufacture is the important factor. The piping systems for BWRs are mostly made of carbon steel, and stainless steel pipings are used for the recirculation system cooling reactors and instrumentation system. Recently, grain boundary type stress corrosion cracking has occurred in the heat-affected zones of welded stainless steel pipings in some BWR plants. In this paper, the quality control of stainless steel pipings is described from the standpoint of preventing stress corrosion cracking in BWR plants. The pipings for nuclear power plants must have sufficient toughness so that the sudden rupture never occurs, and also sufficient corrosion resistance so that corrosion products do not raise the radioactivity level in reactors. The stress corrosion cracking occurred in SUS 304 pipings, the factors affecting the quality of stainless steel pipings, the working method which improves the corrosion resistance and welding control are explained. (Kako, I.)

Effect of small-scale biomass gasification at the state of refractory lining the fixed bed reactor

Energy Technology Data Exchange (ETDEWEB)

Janša, Jan, E-mail: jan.jansa@vsb.cz; Peer, Vaclav, E-mail: vaclav.peer@vsb.cz; Pavloková, Petra, E-mail: petra.pavlokova@vsb.cz [VŠB – Technical University of Ostrava, Energy Research Center, 708 33 Ostrava (Czech Republic)

2016-06-30

The article deals with the influence of biomass gasification on the condition of the refractory lining of a fixed bed reactor. The refractory lining of the gasifier is one part of the device, which significantly affects the operational reliability and durability. After removing the refractory lining of the gasifier from the experimental reactor, there was done an assessment how gasification of different kinds of biomass reflected on its condition in terms of the main factors affecting its life. Gasification of biomass is reflected on the lining, especially through sticking at the bottom of the reactor. Measures for prolonging the life of lining consist in the reduction of temperature in the reactor, in this case, in order to avoid ash fusion biomass which it is difficult for this type of gasifier.

On-line test of power distribution prediction system for boiling water reactors

International Nuclear Information System (INIS)

Nishizawa, Y.; Kiguchi, T.; Kobayashi, S.; Takumi, K.; Tanaka, H.; Tsutsumi, R.; Yokomi, M.

1982-01-01

A power distribution prediction system for boiling water reactors has been developed and its on-line performance test has proceeded at an operating commercial reactor. This system predicts the power distribution or thermal margin in advance of control rod operations and core flow rate change. This system consists of an on-line computer system, an operator's console with a color cathode-ray tube, and plant data input devices. The main functions of this system are present power distribution monitoring, power distribution prediction, and power-up trajectory prediction. The calculation method is based on a simplified nuclear thermal-hydraulic calculation, which is combined with a method of model identification to the actual reactor core state. It has been ascertained by the on-line test that the predicted power distribution (readings of traversing in-core probe) agrees with the measured data within 6% root-mean-square. The computing time required for one prediction calculation step is less than or equal to 1.5 min by an HIDIC-80 on-line computer

Resolution of thermal striping issue downstream of a horizontal pipe elbow in stratified pipe flow. [LMFBR

Energy Technology Data Exchange (ETDEWEB)

Kuzay, T.M.; Kasza, K.E.

1985-01-01

A thermally stratified pipe flow produced by a thermal transient when passing through a horizontal elbow as a result of secondary flow gives rise to large thermal fluctuations on the inner curvature wall of the downstream piping. These fluctuations were measured in a specially instrumented horizontal pipe and elbow system on a test set-up using water in the Mixing Components Technology Facility (MCTF) at Argonne National Laboratory (ANL). This study is part of a larger program which is studying the influence of thermal buoyancy on general reactor component performance. This paper discusses the influence of pipe flow generated thermal oscillations on the thermal stresses induced in the pipe walls. The instrumentation was concentrated around the exit plane of the 90/sup 0/ sweep elbow, since prior tests had indicated that the largest thermal fluctuations would occur within about one hydraulic diameter downstream of the elbow exit. The thermocouples were located along the inner curvature of the piping and measured the near surface fluid temperature. The test matrix involved thermal downramps under turbulent flow conditions.

TARMS, an on-line boiling water reactor operation management system

International Nuclear Information System (INIS)

Iwamoto, T.; Sakurai, S.; Uematsu, H.; Tsuiki, M.; Makino, K.

1984-01-01

The TARMS (Toshiba Advanced Reactor Management System) software package was developed as an effective on-line, on-site tool for boiling water reactor core operation management. It was designed to support a complete function set to meet the requirement to the current on-line process computers. The functions can be divided into two categories. One is monitoring of the present core power distribution as well as related limiting parameters. The other is aiding site engineers or reactor operators in making the future reactor operating plan. TARMS performs these functions with a three-dimensional BWR core physics simulator LOGOS 2, which is based on modified one-group, coarse-mesh nodal diffusion theory. A method was developed to obtain highly accurate nodal powers by coupling LOGOS 2 calculations with the readings of an in-core neutron flux monitor. A sort of automated machine-learning method also was developed to minimize the errors caused by insufficiency of the physics model adopted in LOGOS 2. In addition to these fundamental calculational methods, a number of core operation planning aid packages were developed and installed in TARMS, which were designed to make the operator's inputs simple and easy. (orig.) [de

49 CFR 192.321 - Installation of plastic pipe.

Science.gov (United States)

2010-10-01

... 49 Transportation 3 2010-10-01 2010-10-01 false Installation of plastic pipe. 192.321 Section 192... Transmission Lines and Mains § 192.321 Installation of plastic pipe. (a) Plastic pipe must be installed below ground level except as provided by paragraphs (g) and (h) of this section. (b) Plastic pipe that is...

49 CFR 192.125 - Design of copper pipe.

Science.gov (United States)

2010-10-01

... 49 Transportation 3 2010-10-01 2010-10-01 false Design of copper pipe. 192.125 Section 192.125... BY PIPELINE: MINIMUM FEDERAL SAFETY STANDARDS Pipe Design § 192.125 Design of copper pipe. (a) Copper... hard drawn. (b) Copper pipe used in service lines must have wall thickness not less than that indicated...

Mechanical properties of roll extruded nuclear reactor piping

International Nuclear Information System (INIS)

Steichen, J.M.; Knecht, R.L.

1975-01-01

The elevated temperature mechanical properties of large diameter (28 inches) seamless pipe produced by roll extrusion for use as primary piping for sodium coolant in the Fast Flux Test Facility (FFTF) have been characterized. The three heats of Type 316H stainless steel piping material used exhibited consistent mechanical properties and chemical compositions. Tensile and creep-rupture properties exceeded values on which the allowable stresses for ASME Code Case 1592 on Nuclear Components in Elevated Temperature Service were based. Tensile strength and ductility were essentially unchanged by aging in static sodium at 1050 0 F for times to 10,000 hours. High strain rate tensile tests showed that tensile properties were insensitive to strain rate at temperatures to 900 0 F and that for temperatures of 1050 0 F and above both strength and ductility significantly increased with increasing strain rate. Fatigue-crack propagation properties were comparable to results obtained on plate material and no differences in crack propagation were found between axial and circumferential orientations. (U.S.)

Stabilization of magnet assemblies of permanent magnet sodium flowmeters used in fast breeder reactors

Energy Technology Data Exchange (ETDEWEB)

Rajan, K.K., E-mail: kkrajan@igcar.gov.in; Vijayakumar, G.

2014-08-15

Highlights: • Stabilization procedure for ALNICO-5 permanent magnet material is evolved. • Effect of time and temperature on ALNICO-5 assembly is determined. • Suitability of ALNICO-5 flowmeters at high temperatures is established. • Temperature coefficient of flux density is determined. - Abstract: Permanent magnet flow meters (PMFMs) are used to measure the sodium flow in sodium cooled Fast Breeder Reactor Circuits. Prototype fast breeder reactor (PFBR) which is under construction at Kalpakkam is a 500 MWe, sodium cooled, pool type reactor. Sodium flow measurement in various loops of the reactor is of prime importance from operational and safety point of view. To measure the flow of electrically conducting liquid sodium, in primary and secondary circuit pipe lines of PFBR, permanent magnet flow meters are used. PMFM is a non-invasive device, which works on the principle of generation of motional EMF by magnetic forces exerted on the charges in a moving conductor. Flowmeters of different pipe sizes ranging from 10 mm to 200 mm pipe diameter are required for PFBR. Long term performance of the flowmeters mainly depends on stability of permanent magnets used in flowmeters to generate constant magnetic field in stainless steel (SS) pipes. This paper describes the effects of time and temperature on permanent magnet assemblies made of ALNICO-V used in PFBR flowmeters. The stabilization methodology for ALNICO-V permanent magnet assemblies is evolved and established. Loss of magnetic field strength with respect to time and temperatures is determined by experiments and found negligible.

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.)

Report of the US Nuclear Regulatory Commission Piping Review Committee. Volume 5. Summary - Piping Review Committee conclusions and recommendations

International Nuclear Information System (INIS)

1985-04-01

This document summarizes a comprehensive review of NRC requirements for Nuclear Piping by the US NRC Piping Review Committee. Four topical areas, addressed in greater detail in Volumes 1 through 4 of this report, are included: (1) Stress Corrosion Cracking in Piping of Boiling Water Reactor Plants; (2) Evaluation of Seismic Design; (3) Evaluation of Potential for Pipe Breaks; and (4) Evaluation of Other Dynamic Loads and Load Combinations. This volume summarizes the major issues, reviews the interfaces, and presents the Committee's conclusions and recommendations for updating NRC requirements on these issues. This report also suggests research or other work that may be required to respond to issues not amenable to resolution at this time

49 CFR 179.400-17 - Inner tank piping.

Science.gov (United States)

2010-10-01

... connected to this line to operate at their design capacity without excessive pressure build-up in the tank... housing and must be directed upward and away from operating personnel. (b) Any pressure building system...-17 Inner tank piping. (a) Product lines. The piping system for vapor and liquid phase transfer and...

Piping data retrieval system (PDRS): An integrated package to aid piping layout

International Nuclear Information System (INIS)

Vyas, K.N.; Sharma, A.; Susandhi, R.; Basu, S.

1986-01-01

An integrated package to aid piping layout has been developed and implemented on PDP-11/34 system at Hall 7. The package allows various equipments to be modelled, consisting of primitive equipment components. The equipment layout for the plant can then be reproduced in the form of drawings such as plan, elevation, isometric or perspective. The package has the built in function to perform hidden line removal among equipments. Once the equipment layout is finalised, the package aids in superimposing the piping as per the specified pipe routine. The report discusses the general capabilities and the major input requirements for the package. (author)

Apparatus for measuring total flow in pipes

International Nuclear Information System (INIS)

Matthews, H.

1986-01-01

To obtain a sample representative of the total flow in a pipe over a given period a Pitot tube is located in the pipe and connected to a collector outside the pipe. The collector is pressurised to a pressure substantially equal to the static head of the flow in the pipe via a line. Liquid is discharged from a collector to a container which is vented to atmosphere. (author)

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

Impacting effects of seismic loading in feeder pipes of PHWR power plants

International Nuclear Information System (INIS)

Kumar, R.

1996-01-01

The core of a pressurized heavy water reactor (PHWR) consists of a large number of fuel channels. These fuel channels are connected to the feeder pipes through which the heavy water flows and transports heat from the reactor core to the steam generators. The feeder pipes are several hundreds in number. They run close to each other with small gaps and have several bends. Thus they represent a complex piping system. Under seismic loading, the adjacent feeder pipes may impact each other. In this paper a simplified procedure has been established to assess such impacting effects. The results of the proposed analysis include bending moment and impact force, which provide the stresses due to impacting effects. These results are plotted in nondimensional form so that they could be utilized for any set of feeder pipes. The procedure used for studying the impacting effects includes seismic analysis of individual feeder pipes without impacting effects, selection of pipes for impact analysis, and estimating their maximum impact velocity. Based on the static and dynamic characteristics of the selected feeder pipes, the maximum bending moment, impact force, and stresses are obtained. The results of this study are useful for quick evaluation of the impacting effects in feeder pipes

Pipe line systems in nuclear power plant

International Nuclear Information System (INIS)

Sasada, Yasuhiro; Tanno, Kazuo; Shibato, Eizo.

1979-01-01

Purpose: To prevent stress corrosion cracks, in particular, for branched pipeways by conducting water quality control in the branched pipeways as well as in the main pipeways, and reducing the thermal stress in the branched pipeways. Constitution: A water quality monitoring device is provided to a drain pipe and a failed element detection pipe to monitor the quality of stagnated water continuously or periodically. If the impurity concentration or oxygen concentration exceeds a specified value in the stagnated water, a drain valve or a check valve is opened by a signal from the water quality monitoring device to replace the stagnated water with recycling water in the main pipeway. The temperature for the branched loop pipeway and the main pipeway are collectively kept to a same temperature to thereby reduce the thermal stress in the branched pipeway. (Kawakami, Y.)

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

Simulation experiments of small break LOCA in upper plenum joint pipe for 5 MW heating test reactor

International Nuclear Information System (INIS)

Bo Jinhai; Jiang Shengyao; Zhang Youjie; Tong Yunxian; Sun Shusen; Yao Meisheng

1988-12-01

A simulation experiment of small break LOCA is introduced, which was caused by the breakdown of a small size or middle size pipe located at upper plenum, or by unexpected opening the safety valve. In the tests, the system pressure, temperature, void fraction and total loss of water were studied. The results showed that the total loss of water was nearly 20% of initial loading water. It means under this condition the 5MW low temperature heating reactor being built in Institute of Nnclear Engergy Technology of Tsinghua University is safe

Space reactor preliminary mechanical design

International Nuclear Information System (INIS)

Meier, K.L.

1983-01-01

An analysis was performed on the SABRE reactor space power system to determine the effect of the number and size of heat pipes on the design parameters of the nuclear subsystem. Small numbers of thin walled heat pipes were found to give a lower subsystem mass, but excessive fuel swelling resulted. The SP-100 preliminary design uses 120 heat pipes because of acceptable fuel swelling and a minimum nuclear subsystem mass of 1875 kg. Salient features of the reactor preliminary design are: individual fuel modules, ZrO 2 block core mounts, bolted collar fuel module restraints, and a BeO central plug

IEA-R1 primary and secondary coolant piping systems coupled stress analysis

International Nuclear Information System (INIS)

Fainer, Gerson; Faloppa, Altair A.; Oliveira, Carlos A.; Mattar Neto, Miguel

2013-01-01

The aim of this work is to perform the stress analysis of a coupled primary and secondary piping system of the IEA-R1 based on tridimensional model, taking into account the as built conditions. The nuclear research reactor IEA-R1 is a pool type reactor projected by Babcox-Willcox, which is operated by IPEN since 1957. The operation to 5 MW power limit was only possible after the conduction of life management and modernization programs in the last two decades. In these programs the components of the coolant systems, which are responsible for the water circulation into the reactor core to remove the heat generated inside it, were almost totally refurbished. The changes in the primary and secondary systems, mainly the replacement of pump and heat-exchanger, implied in piping layout modifications, and, therefore, the stress condition of the piping systems had to be reanalyzed. In this paper the structural stress assessment of the coupled primary and secondary piping systems is presented and the final results are discussed. (author)

Comparative study of computational model for pipe whip analysis

International Nuclear Information System (INIS)

Koh, Sugoong; Lee, Young-Shin

1993-01-01

Many types of pipe whip restraints are installed to protect the structural components from the anticipated pipe whip phenomena of high energy lines in nuclear power plants. It is necessary to investigate these phenomena accurately in order to evaluate the acceptability of the pipe whip restraint design. Various research programs have been conducted in many countries to develop analytical methods and to verify the validity of the methods. In this study, various calculational models in ANSYS code and in ADLPIPE code, the general purpose finite element computer programs, were used to simulate the postulated pipe whips to obtain impact loads and the calculated results were compared with the specific experimental results from the sample pipe whip test for the U-shaped pipe whip restraints. Some calculational models, having the spring element between the pipe whip restraint and the pipe line, give reasonably good transient responses of the restraint forces compared with the experimental results, and could be useful in evaluating the acceptability of the pipe whip restraint design. (author)

Emergency core cooling system for a fast reactor

International Nuclear Information System (INIS)

Johnson, H.G.; Madsen, R.N.

1976-01-01

The main heat transport system for a liquid-metal-cooled nuclear reactor is constructed with elevated piping and guard vessels or pipes around all components of the system below the elevation of the elevated piping so the head developed by the pumps at emergency motor speed will be unsufficient to lift the liquid-metal-coolant over the top of the guard tanks or pipes or out of the elevated piping in the event of a loss-of-coolant accident. In addition, inlet downcomers to the reactor vessel are contained within guard standpipes having a clearance volume as small as practicable. 4 claims, 2 drawing figures

Assessment of LWR piping design loading based on plant operating experience

International Nuclear Information System (INIS)

Svensson, P.O.

1980-08-01

The objective of this study has been to: (1) identify current Light Water Reactor (LWR) piping design load parameters, (2) identify significant actual LWR piping loads from plant operating experience, (3) perform a comparison of these two sets of data and determine the significance of any differences, and (4) make an evaluation of the load representation in current LWR piping design practice, in view of plant operating experience with respect to piping behavior and response to loading

Development of an advanced PFM code for the integrity evaluation of nuclear piping system under combined aging mechanisms

International Nuclear Information System (INIS)

Datta, Debashis

2010-02-01

A nuclear piping system is composed of several straight pipes and elbows joined by welding. These weld sections are usually the most susceptible failure parts susceptible to various degradation mechanisms. Whereas a specific location of a reactor piping system might fail by a combination of different aging mechanisms, e.g. fatigue and/or stress corrosion cracking, the majority of the piping probabilistic fracture mechanics (PFM) codes can only consider a single aging mechanism at a time. So, a probabilistic fracture mechanics computer code capable of considering multiple aging mechanisms was developed for an accurate failure analysis of each specific component of a nuclear piping section. The newly proposed crack morphology based probabilistic leak flow rate module is introduced in this code to separately treat fatigue and SCC type cracks. Improved models e.g. stressors models, elbow failure model, SIFs model, local seismic occurrence probability model, performance based crack detection models, etc., are also included in this code. Recent probabilistic fatigue (S-N) and SCC crack initiation (S-T) and subsequent crack growth rate models are coded. An integrated probabilistic risk assessment and probabilistic fracture mechanics methodology is proposed. A complete flow chart regarding the combined aging mechanism model is presented. The combined aging mechanism based module can significantly reduce simulation efforts and time. Two NUREG benchmark problems, e.g. reactor pressure vessel outlet nozzle section and a surge line elbow located just below the pressurizer are reinvestigated by this code. The results showed that, contribution of pre-existing cracks in addition to initiating cracks, can significantly increase the overall failure probability. Inconel weld location of reactor pressure vessel outlet nozzle section showed the weakest point in terms of relative through-wall leak failure probability in the order of about 10 -2 at the 40-year plant life. Considering

View of industry on the impact of pipe break criteria

International Nuclear Information System (INIS)

Bernsen, S.A.

1983-01-01

Historically, large pipe breaks in the types of materials used and under operating conditions similar to those in light water reactor service have not occurred. Nevertheless, the non-mechanistic assumption of a double ended pipe break of the early sixties, selected for loss of coolant accident analysis purposes, has become a mechanistic criterion for the design and arrangement of high pressure piping systems and their associated supports and enclosures in today's nuclear plants. While it seems reasonable and appropriate to continue to design the Emergency Core Cooling Systems for a range of loss of coolant accidents up to and including those that approximate the area of the largest pipe connected to the reactor vessel and to use this break in determining the loading and temperature rise rate for containment structures and equipment qualification, it no longer seems reasonable to provide precisely engineered break protection for a limited number of potential pipe break locations. This observation is gaining increasing support, particularly as engineering judgment and historical perspectives are being supplemented by both deterministic and probabilistic studies that indicate the potential for large instantaneous breaks in nuclear grade piping systems is virtually incredible. Fracture mechanics analyses support leak-before-break assumptions with wide margins and probabilistic studies indicate potentials for double-ended pipe breaks in the range of less than one in a billion years

Ductile fracture behaviour of primary heat transport piping material ...

Indian Academy of Sciences (India)

R. Narasimhan (Krishtel eMaging) 1461 1996 Oct 15 13:05:22

Abstract. Design of primary heat transport (PHT) piping of pressurised heavy water reactors (PHWR) has to ensure implementation of leak-before-break con- cepts. In order to be able to do so, the ductile fracture characteristics of PHT piping material have to be quantified. In this paper, the fracture resistance of SA333, Grade.

Application of tearing modulus stability concepts to nuclear piping. Final report. [PWR; BWR

Energy Technology Data Exchange (ETDEWEB)

Cotter, K.H.; Chang, H.Y.; Zahoor, A.

1982-02-01

The recently developed tearing modulus stability concept was successfully applied to several boiling water reactor (BWR) and pressurized water reactor (PWR) piping systems. Circumferentially oriented through-the-thickness cracks were postulated at numerous locations in each system. For each location, the simplified tearing stability methods developed in USNRC Report NUREG/CR-0838 were used to determine crack stability. The J-T diagram was used to present the results of the computations. The piping systems considered included Type 304 stainless steel as well as A106 carbon steel materials. These systems were analyzed using the piping analysis computer code MINK.

Selection of power plant elements for future reactor space electric power systems

International Nuclear Information System (INIS)

Buden, D.; Bennett, G.A.; Copper, K.

1979-09-01

Various types of reactor designs, electric power conversion equipment, and reject-heat systems to be used in nuclear reactor power plants for future space missions were studied. The designs included gas-cooled, liquid-cooled, and heat-pipe reactors. For the power converters, passive types such as thermoelectric and thermionic converters and dynamic types such as Brayton, potassium Rankine, and Stirling cycles were considered. For the radiators, heat pipes for transfer and radiating surface, pumped fluid for heat transfer with fins as the radiating surface, and pumped fluid for heat transfer with heat pipes as the radiating surface were considered. After careful consideration of weights, sizes, reliabilities, safety, and development cost and time, a heat-pipe reactor design, thermoelectric converters, and a heat-pipe radiator for an experimental program were selected

Test results of a jet impingement from a 4 inch pipe under BWR LOCA conditions

International Nuclear Information System (INIS)

Isozaki, Toshikuni; Yano, Toshikazu; Miyazaki, Noriyuki; Kato, Rokuro; Kurihara, Ryoichi; Ueda, Shuzo; Miyazono, Shohachiro

1982-09-01

Hypothetical instantaneous pipe rupture is now considered to be one of the design basis accidents during the operation of the light water reactor. If a pipe rupture accidnet occurs, the pipe will start moving with the sudden discharge of internal fluid. So, the various apparatus such as pipe whip restraints and jet deflectors are being installed near the postulated break location to protect the nuclear power plants against the effect of postulated pipe rupture. Pipe whipping test and jet discharge test are now being conducted at the Division of Reactor Safety of the Japan Atomic Energy Research Institute. This report describes the test results of the jet discharge from a 4 inch pipe under BWR LOCA condition. In front of the pipe exit the target disk of 1000 mm in diameter was installed. The distance between the pipe exit and the target was 500 mm. 13 pressure transducers and 13 thermocouples were mounted on the target disk to measure the pressure and temperature increase due to jet impingement on the target. (author)

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.)

A method to assign failure rates for piping reliability assessments

International Nuclear Information System (INIS)

Gamble, R.M.; Tagart, S.W. Jr.

1991-01-01

This paper reports on a simplified method that has been developed to assign failure rates that can be used in reliability and risk studies of piping. The method can be applied on a line-by-line basis by identifying line and location specific attributes that can lead to piping unreliability from in-service degradation mechanisms and random events. A survey of service experience for nuclear piping reliability also was performed. The data from this survey provides a basis for identifying in-service failure attributes and assigning failure rates for risk and reliability studies

R and D of On-line Reprocessing Technology for Molten-Salt Reactor Systems

International Nuclear Information System (INIS)

Uhlir, Jan; Tulackova, Radka; Chuchvalcova Bimova, Karolina

2006-01-01

The Molten Salt Reactor (MSR) represents one of promising future nuclear reactor concept included in the Generation IV reactors family. The reactor can be operated as the thorium breeder or as the actinide transmuter. However, the future deployment of Molten-Salt Reactors will be significantly dependent on the successful mastering of advanced reprocessing technologies dedicated to their fuel cycle. Here the on-line reprocessing technology connected with the fuel circuit of MSR is of special importance because the reactor cannot be operated for a long run without the fuel salt clean-up. Generally, main MSR reprocessing technologies are pyrochemical, majority of them are fluoride technologies. The proposed flow-sheets of MSR on-line reprocessing are based on a combination of molten-salt / liquid metal extraction and electro-separation processes, which can be added to the gas extraction process already verified during the MSRE project in ORNL. The crucial separation method proposed for partitioning of actinides from fission products is based on successive Anodic dissolution and Cathodic deposition processes in molten fluoride media. (authors)

77 FR 39695 - HollyFrontier Refining and Marketing LLC v. Osage Pipe Line Company, LLC; Notice of Complaint

Science.gov (United States)

2012-07-05

... Refining and Marketing LLC v. Osage Pipe Line Company, LLC; Notice of Complaint Take notice that on June 25...; 18 CFR 343.1(a) and 343.2(c), HollyFrontier Refining and Marketing LLC (Complainant) filed a formal... assistance with any FERC Online service, please email [email protected] , or call (866) 208-3676...

Fabrication and evaluation of chemically vapor deposited tungsten heat pipe.

Science.gov (United States)

Bacigalupi, R. J.

1972-01-01

A network of lithium-filled tungsten heat pipes is being considered as a method of heat extraction from high temperature nuclear reactors. The need for material purity and shape versatility in these applications dictates the use of chemically vapor deposited (CVD) tungsten. Adaptability of CVD tungsten to complex heat pipe designs is shown. Deposition and welding techniques are described. Operation of two lithium-filled CVD tungsten heat pipes above 1800 K is discussed.

Visualization of crust in metallic piping through real-time neutron radiography obtained with low intensity thermal neutron flux

Energy Technology Data Exchange (ETDEWEB)

Luiz, Leandro C.; Crispim, Verginia R. [Nuclear Engineering Program, Federal University of Rio de Janeiro, Rio de Janeiro (Brazil); Ferreira, Francisco J. O. [National Nuclear Energy Commission, CNEN/IEN, Division Reactors, Rio de Janeiro (Brazil)

2017-06-15

The presence of crust on the inner walls of metallic ducts impairs transportation because crust completely or partially hinders the passage of fluid to the processing unit and causes damage to equipment connected to the production line. Its localization is crucial. With the development of the electronic imaging system installed at the Argonauta/Nuclear Engineering Institute (IEN)/National Nuclear Energy Commission (CNEN) reactor, it became possible to visualize crust in the interior of metallic piping of small diameter using real-time neutron radiography images obtained with a low neutron flux. The obtained images showed the resistance offered by crust on the passage of water inside the pipe. No discrepancy of the flow profile at the bottom of the pipe, before the crust region, was registered. However, after the passage of liquid through the pipe, images of the disturbances of the flow were clear and discrepancies in the flow profile were steep. This shows that this technique added the assembled apparatus was efficient for the visualization of the crust and of the two-phase flows.

Visualization of crust in metallic piping through real-time neutron radiography obtained with low intensity thermal neutron flux

International Nuclear Information System (INIS)

Luiz, Leandro C.; Crispim, Verginia R.; Ferreira, Francisco J. O.

2017-01-01

The presence of crust on the inner walls of metallic ducts impairs transportation because crust completely or partially hinders the passage of fluid to the processing unit and causes damage to equipment connected to the production line. Its localization is crucial. With the development of the electronic imaging system installed at the Argonauta/Nuclear Engineering Institute (IEN)/National Nuclear Energy Commission (CNEN) reactor, it became possible to visualize crust in the interior of metallic piping of small diameter using real-time neutron radiography images obtained with a low neutron flux. The obtained images showed the resistance offered by crust on the passage of water inside the pipe. No discrepancy of the flow profile at the bottom of the pipe, before the crust region, was registered. However, after the passage of liquid through the pipe, images of the disturbances of the flow were clear and discrepancies in the flow profile were steep. This shows that this technique added the assembled apparatus was efficient for the visualization of the crust and of the two-phase flows

The First Assembly Line of Large-longitudinally-welded Steel Pipe in China Went into Operation

Institute of Scientific and Technical Information of China (English)

Li Bing

2002-01-01

@@ On July 27, the first assembly line to produce JCOE large diameter Longitudinally-submerged-arc-welded steel pipe in China, Which is the key homemade equipment project of "West-East Gas Transmission"project, was put into production. Chen Gen, vice general manager of CNPC; Xie Zhiqiang and Liu Haisheng, assistant chief manager of CNPC; Shi Xingquan, vice president of PetroChina; and the president of Itochu-Marubeni Steel & iron Co., Ltd.of Japan; attended the opening ceremony and cut the ribbon.

TransCanada PipeLines Limited 1996 annual report : profitable today, prepared for tomorrow

International Nuclear Information System (INIS)

1997-01-01

Operations and financial activities of TransCanada PipeLines Limited (TPL) during 1996 were reviewed and made available for the benefit of shareholders. TransCanada PipeLines Limited is a Canadian company with assets in excess of $12 billion. Its pipeline system transports natural gas and crude oil from the Western Canada Sedimentary Basin to North America's major energy markets. By all accounts the company had a successful year in 1996, having delivered a record 2.4 Bcf of gas, up 3.7 per cent from 1995, through the Canadian Mainline, TransCanada's largest asset. Among other notable achievements was the decision to join the Portland Natural Gas Transmission System as a 20 per cent partner in a project to construct a natural gas pipeline between Quebec and Haverhill, Massachusetts. Another accomplishment was the completion of the acquisition of all remaining shares of Alberta Natural Gas company, thereby extending the company's energy transmission, gas processing and specialty chemical operations. Other achievements were the purchase of Enron Louisiana Energy Company, the acquisition of assets that position TPL as one of the largest natural gas liquids processors in North America and the completion and commissioning of two new independent power generation plants in Ontario. Shareholder's return on investment was 34 per cent, more than double the target of between 12 to 15 per cent. Net income to common shares amounted to $1.85 per share, up from $1.75 in 1995. The strong performance was attributed to continued growth in all four segments of TPL's business, i. e. energy transmission, energy marketing, energy processing and international. tabs., figs

Critical element development of standard pipe connector for remote handling

International Nuclear Information System (INIS)

Taguchi, Kou; Kakudate, Satoshi; Kanamori, Naokazu; Oka, Kiyoshi; Nakahira, Masataka; Obara, Kenjiro; Tada, Eisuke; Shibanuma, Kiyoshi; Seki, Masahiro

1994-08-01

In fusion experimental reactors such as ITER, the in-vessel components such as blanket and divertor are actively cooled and a large number of cooling pipes are located around the core of reactor, where personnel access is prohibited. Mechanical pipe connectors are highly required as standard components for easy and reliable connection/disconnection of cooling pipe by remote handling. For this purpose, a clamping chain type connector has been developed with special mechanisms such as plate springs and guide structures so as to enable concentric and axial movement of clamping chain for easy mounting and dismounting. The basic performance test of a prototypical connector for a 80-A pipe shows sufficient leak tightness and proof pressure capability as well as simple connection/disconnection operation. In addition to the clamp chain type connector, design efforts have been made to develop a quick coupling type connector and a preliminary model of air-actuated quick connector has been fabricated for further investigations. This paper gives the design concept of mechanical pipe connectors such as clamping chain type and quick coupler type, and the basic performance tests results of clamping chain type connector. (author)

Evolution of thermal fatigue management of piping in US LWRs

International Nuclear Information System (INIS)

McDewitt, M.; Wolfe, K.; McGill, R.

2015-01-01

Fatigue usage caused by cyclic changes of thermally stratified reactor coolant in Light Water Reactor (LWR) pressure boundary piping was not an original consideration in US Nuclear Power Plant (NPP) designs. During the mid 1980's, several events involving cracking and leakage due to thermal cycling occurred in reactor coolant system branch piping at both US and International NPPs. In 1988, the US Nuclear Regulatory Commission (US NRC) issued Bulletin 88-08 to alert LWR licensees of the potential for piping failures due to stratified thermal cycling. In response to these events, the US nuclear industry developed initiatives to identify susceptible components and established measures to monitor and prevent future failures. These initiatives have been effective in preventing leakage events, but have also identified fewer defects than expected based on screening model predictions. Improved analytical techniques are being investigated to maintain program effectiveness while minimizing unnecessary non-destructive examinations. This paper discusses the evolution of the US thermal fatigue initiatives, and analytical concepts being evaluated to improve program efficiency. (authors)

Preliminary review of mass transfer and flow visualization studies and techniques relevant to the study of erosion-corrosion of reactor piping systems

Energy Technology Data Exchange (ETDEWEB)

Kuzay, T.M.; Halle, H.J.; Kasza, K.E.

1988-06-01

This report provides some background information on the failed piping at the Surry-2 reactor; a summary of pertinent literature on mass transfer in related geometries; and a description of methodologies for visualization and erosion rate measurements in laboratory model studies that can provide greater insight into the role of flow geometry in erosion-corrosion. 18 refs., 9 figs., 1 tab.

Preliminary review of mass transfer and flow visualization studies and techniques relevant to the study of erosion-corrosion of reactor piping systems

International Nuclear Information System (INIS)

Kuzay, T.M.; Halle, H.J.; Kasza, K.E.

1988-06-01

This report provides some background information on the failed piping at the Surry-2 reactor; a summary of pertinent literature on mass transfer in related geometries; and a description of methodologies for visualization and erosion rate measurements in laboratory model studies that can provide greater insight into the role of flow geometry in erosion-corrosion. 18 refs., 9 figs., 1 tab

Self-operation type power control device for nuclear reactor

Energy Technology Data Exchange (ETDEWEB)

Kanbe, Mitsuru.

1993-07-23

The device of the present invention operates by sensing the temperature change of a reactor core in all of LMFBR type reactors irrespective of the scale of the reactor core power. That is, a region where liquid poison is filled is disposed at the upper portion and a region where sealed gases are filled is disposed at the lower portion of a pipe having both ends thereof being closed. When the pipe is inserted into the reactor core, the inner diameter of the pipe is determined smaller than a predetermined value so that the boundary between the liquid poison and the sealed gases in the pipe is maintained relative to an assumed maximum acceleration. The sealed gas region is disposed at the reactor core region. If the liquid poison is expanded by the elevation of the reactor core exit temperature, it is moved to the lower gas region, to control the reactor power. Since high reliability can be maintained over a long period of time by this method, it is suitable to FBR reactors disposed in such environments that maintenance can not easily be conducted, such as desserts, isolated islands and undeveloped countries. Further, it is also suitable to ultra small sized nuclear reactors disposed at environments that the direction and the magnitude of gravity are different from those on the ground. (I.S.).

Self-operation type power control device for nuclear reactor

International Nuclear Information System (INIS)

Kanbe, Mitsuru.

1993-01-01

The device of the present invention operates by sensing the temperature change of a reactor core in all of LMFBR type reactors irrespective of the scale of the reactor core power. That is, a region where liquid poison is filled is disposed at the upper portion and a region where sealed gases are filled is disposed at the lower portion of a pipe having both ends thereof being closed. When the pipe is inserted into the reactor core, the inner diameter of the pipe is determined smaller than a predetermined value so that the boundary between the liquid poison and the sealed gases in the pipe is maintained relative to an assumed maximum acceleration. The sealed gas region is disposed at the reactor core region. If the liquid poison is expanded by the elevation of the reactor core exit temperature, it is moved to the lower gas region, to control the reactor power. Since high reliability can be maintained over a long period of time by this method, it is suitable to FBR reactors disposed in such environments that maintenance can not easily be conducted, such as desserts, isolated islands and undeveloped countries. Further, it is also suitable to ultra small sized nuclear reactors disposed at environments that the direction and the magnitude of gravity are different from those on the ground. (I.S.)

46 CFR 182.455 - Fuel piping.

Science.gov (United States)

2010-10-01

... system is of nickel-copper or copper-nickel. When making tube connections, the tubing must be cut square...) MACHINERY INSTALLATION Specific Machinery Requirements § 182.455 Fuel piping. (a) Materials and workmanship. The materials and construction of fuel lines, including pipe, tube, and hose, must comply with the...

Status of FRJ-2 refurbishment of tank pipes and essential results of aging analysis

International Nuclear Information System (INIS)

Hansen, G.; Thamm, G.; Thome, M.

1993-01-01

An aging evaluation program for FRJ-2 (DIDO) of the Forschungszentrum Juelich GmbH has been developed and is currently executed in cooperation with the licensing and regulatory and TUV experts in order to determine the overall life expectancy of the facility and to identify critical systems and components that need to be upgraded or refurbished for future safe reactor operation. In Phase A (completed) a so called master list of the FRJ-2 mechanical, electrical and structural components was compiled on a system-by system basis and the operational documentation with respect to regular inspections, maintenance, repair and unusual occurences was carefully examined. Critical components were selected and their ageing respectively life limiting mechanisms identified. In Phase (currently under way) special inspections, examinations and tests for critical systems/components are being elaborated, executed and evaluated. Current work is being concentrated on non replaceable components (e.g. reactor aluminium tank (RAT) and the connecting pipes to the primary cooling circuit, the reactor steel tank and pipe work inside the concrete reactor block). As a consequence of first results of the aging evaluation program and due to leaks in the weir and drain pipes of the RAT a repair/refurbishment program was set up for the Al-RAT pipes (risers, downcomers weir and drain pipes) and the steel guide tubes. Details of the r/r program which is in far progress and first essential results of the aging evaluation will be presented. The results achieved until today are encouraging with respect to safe reactor operation on short and medium term. (author)

Status of FRJ-2 Refurbishment of tank pipes and essential results of aging analysis

International Nuclear Information System (INIS)

Hansen, G.; Thamm, G.; Thome, M.

1994-01-01

An aging evaluation program for FRJ-2 (DIDO) of the Forschungszentrum Juelich GmbH has been developed and is currently executed in cooperation with the licensing and regulatory and TUEV experts in order to determine the overall life expectancy of the facility and to identify critical systems and components that need to be upgraded or refurbished for future safe reactor operation. In Phase A (completed) a so called master list of the FRJ-2 mechanical, electrical and structural components was compiled on a system-by system basis and the operational documentation with respect to regular inspections, maintenance, repair and unusual occurrences was carefully examined. Critical components were selected and their ageing respectively life limiting mechanisms identified. In Phase B (currently under way) special inspections, examinations and tests for critical systems/components are being elaborated, executed and evaluated. Current work is being concentrated on non replaceable components (e.g. reactor aluminium tank (RAT) and the connecting pipes to the primary cooling circuit, the reactor steel tank and pipe work inside the concrete reactor block). As a consequence of first results of the aging evaluation program and due to leaks in the weir and drain pipes of the RAT a repair/refurbishment program was set up for the Al-RAT pipes (risers, downcomers, weir and drain pipes) and the steel guide tubes. Details of the r/r program which is in far progress and first essential results of the aging evaluation will be presented. The results achieved until today are encouraging with respect to safe reactor operation on short and medium term. (J.P.N.)

A Study on Temperature Distribution in the Hot Leg Pipes considering the Variation of Flow Rate in RCS

Energy Technology Data Exchange (ETDEWEB)

Cho, Hyuksu; Yi, Kunwoo; Choe, Yoonjae; Jang, Hocheol; Yune, Seokjeong; Park, Seongchan [KEPCO Engineering and Construction, Daejeon (Korea, Republic of)

2016-10-15

In this study, a computational analysis is performed to predict the deviation in the temperature distribution in the hot leg pipe according to the flow rate variation in RCS. In the hot leg pipes of Reactor Coolant System (RCS) of APR1400, four Resistance Temperature Detectors (RTDs), to obtain the average hot leg temperature, are installed at each hot leg pipe (two in the upper region and the other two in the lower region around the wall of the hot leg pipe). There is a deviation in temperature distribution in the hot leg pipe due to the sudden changes in the flow direction and area from the reactor core exit to the hot leg pipe. The non-uniform temperature distribution in the hot leg pipe can affect the measurement of the plant parameters such as the reactor power and the reactor coolant flow rate. The following conclusions are reached 1) The non-uniform temperature distribution in the core exit is sustained to some extent through the entire region of hot leg pipe. 2) The temperature ranges having a uniform pattern are 45 - 120° and 240 - 315°. The sensor positions of RTDs are located in this interval (45 - 120° and 240 - 315°) and this sensor positions of RTDs show the appropriate temperature measurement. Also, the temperature distribution shows the similar pattern without reference to the flow rate variation in RCS.

Computer simulation of LMFBR piping systems

International Nuclear Information System (INIS)

A-Moneim, M.T.; Chang, Y.W.; Fistedis, S.H.

1977-01-01

Integrity of piping systems is one of the main concerns of the safety issues of Liquid Metal Fast Breeder Reactors (LMFBR). Hypothetical core disruptive accidents (HCDA) and water-sodium interaction are two examples of sources of high pressure pulses that endanger the integrity of the heat transport piping systems of LMFBRs. Although plastic wall deformation attenuates pressure peaks so that only pressures slightly higher than the pipe yield pressure propagate along the system, the interaction of these pulses with the different components of the system, such as elbows, valves, heat exchangers, etc.; and with one another produce a complex system of pressure pulses that cause more plastic deformation and perhaps damage to components. A generalized piping component and a tee branching model are described. An optional tube bundle and interior rigid wall simulation model makes such a generalized component model suited for modelling of valves, reducers, expansions, and heat exchangers. The generalized component and the tee branching junction models are combined with the pipe-elbow loop model so that a more general piping system can be analyzed both hydrodynamically and structurally under the effect of simultaneous pressure pulses

Managing the Cost of Plant Piping System Leakage

International Nuclear Information System (INIS)

Jenco, John M.; Keck, Donna R.; Johnson, Gary L.

2002-01-01

Recent studies have shown that the average annual cost impact of external piping system leakage on commercial nuclear plant operations and maintenance can easily range into the millions of dollars for each reactor unit. Evidence suggests that this significant O and M cost reduction opportunity has largely been overlooked, due to the number of diverse line items and budget areas affected. Results released last year from an EPRI pilot study of more than a dozen reactor units at seven plant sites operated by multiple utilities found that the average annual cost impact was indeed around $1.6 million per year per unit. Subsequent field experience has also demonstrated that an effective fluid leak management program can substantially reduce these costs within the first three years of implementation. This paper presents the general cost impact research results from various studies, outlines key elements of an effective plant fluid leak management program, discusses important implementation issues, and presents results from case studies covering different utility approaches to developing and implementing an effective fluid leak management program. Actual cost data will be included where appropriate. (authors)

Mitigating energy loss on distribution lines through the allocation of reactors

Science.gov (United States)

Miranda, T. M.; Romero, F.; Meffe, A.; Castilho Neto, J.; Abe, L. F. T.; Corradi, F. E.

2018-03-01

This paper presents a methodology for automatic reactors allocation on medium voltage distribution lines to reduce energy loss. In Brazil, some feeders are distinguished by their long lengths and very low load, which results in a high influence of the capacitance of the line on the circuit’s performance, requiring compensation through the installation of reactors. The automatic allocation is accomplished using an optimization meta-heuristic called Global Neighbourhood Algorithm. Given a set of reactor models and a circuit, it outputs an optimal solution in terms of reduction of energy loss. The algorithm is also able to verify if the voltage limits determined by the user are not being violated, besides checking for energy quality. The methodology was implemented in a software tool, which can also show the allocation graphically. A simulation with four real feeders is presented in the paper. The obtained results were able to reduce the energy loss significantly, from 50.56%, in the worst case, to 93.10%, in the best case.

Reactor design for nuclear electric propulsion

International Nuclear Information System (INIS)

Koenig, D.R.; Ranken, W.A.

1979-01-01

Conceptual design studies of a nuclear power plant for electric propulsion of spacecrafts have been on going for several years. An attractive concept which has evolved from these studies and which has been described in previous publications, is a heat-pipe cooled, fast spectrum nuclear reactor that provides 3 MW of thermal energy to out-of-core thermionic converters. The primary motivation for using heat pipes is to provide redundancy in the core cooling system that is not available in gas or liquid-metal cooled reactors. Detailed investigation of the consequences of heat pipe failures has resulted in modifications to the basic reactor design and has led to consideration of an entirely different core design. The new design features an integral laminated core configuration consisting of alternating layers of UO 2 and molybdenum sheets that span the entire diameter of the core. Design characteristics are presented and compared for the two reactors

Main steam system piping response under safety/relief valve opening events

International Nuclear Information System (INIS)

Swain, E.O.; Esswein, G.A.; Hwang, H.L.; Nieh, C.T.

1980-01-01

The stresses in the main steam branch pipe of a Boiling Water Reactor due to safety/relief valve blowdown has been measured from an in situ piping system test. The test results were compared with analytical results. The predicted stresses using the current state of art analytical methods used for BWR SRV discharge transient piping response loads were found to be conservative when compared to the measured stress values. 3 refs

A study of inter linkage effects on Candu feeder piping

International Nuclear Information System (INIS)

Li, M.; Aggarwal, M.L.; Meysner, A.

2005-01-01

A CANDU (Canadian Deuterium Uranium) reactor core consists of a large number of fuel channels where heat is generated. Two feeder pipes are connected to each fuel channel to transport D 2 O coolant into and out of the reactor core. The feeder piping is designed to the requirements of Class 1 piping of Section III NB of the ASME Boiler and Pressure Vessel and CSA Codes. Feeder piping stress analysis is being performed to demonstrate the code compliance check and the fitness for service of feeders. In the past, stress analyses were conducted for each individual feeder without including interaction effects among connected feeders. Interaction effects occur as a result of linkages that exist between feeders to prevent fretting and impacting damage during normal, abnormal and accident conditions. In this paper, a 'combined' approach is adopted to include all feeders connected by inter linkages into one feeder piping model. MSC/NASTRAN finite element software was used in the stress simulation, which contains up to 127 feeder pipes. The ASME Class 1 piping analysis was conducted to investigate the effects of the linkages between feeders. Both seismic time history and broadened response spectra methods were used in the seismic stress calculation. The results show that the effect of linkages is significant in dynamic stresses for all feeder configurations, as well as in static stresses for certain feeder configurations. The single feeder analysis could either underestimate or overestimate feeder stresses depending on the pipe geometry and bend wall thickness. (authors)

Neutron streaming evaluation for the DREAM fusion power reactor

International Nuclear Information System (INIS)

Seki, Yasushi; Nishio, Satoshi; Ueda, Shuzo; Kurihara, Ryoichi

2000-01-01

Aiming at high degree of safety and benign environmental effect, we have proposed a tokamak fusion reactor concept called DREAM, which stands for DRastically EAsy Maintenance Reactor. The blanket structure of the reactor is made from very low activation SiC/SiC composites and cooled by non-reactive helium gas. High net thermal efficiency of about 50% is realized by 900 C helium gas and high plant availability is possible with simple maintenance scheme. In the DREAM Reactor, neutron streaming is a big problem because cooling pipes with diameter larger than 80 cm are used for blanket heat removal. Neutron streaming through the cooling pipes could cause hot spots in the superconducting magnets adjacent to the cooling pipes to shorten the magnet lifetime or increase cryogenic cooling requirement. Neutron streaming could also activate components such as gas turbine further away from the fusion plasma. The effect of neutron streaming through the helium cooling pipes was evaluated for the two types of cooling pipe extraction scheme. The result of a preliminary calculation indicates the gas turbine activation prohibits personnel access in the case of inboard pipe extraction while with additional shielding measures, limited contact maintenance is possible in the case of outboard extraction. (author)

Isolation colling device for reactor

International Nuclear Information System (INIS)

Ikehara, Morihiko; Arai, Shigeki.

1982-01-01

Purpose: To prevent undesired operation of an emergency core cooling system due to excess lowering of water level in a reactor. Constitution: In an emergency facility adapted to drive a turbine, upon reactor isolation, with the excess steams of the reactor to operate a pump and thereby inject cooling water to the reactor, a water level detector is provided and connected to a pump exhaust valve control circuit, a turbine inlet valve control circuit and a by-pass valve control circuit. Valve ON-OFF is automatically controlled depending on the water level to thereby render the level constant. A by-pass pipe is branched from a pump exhaust pipe and connected to a condensate storage tank. When the water level rises due to water injection, the injecting water is returned to circulate by way of the by-pass pipe to the condensate storage tank under the ON-OFF for each of the valves while the turbine being kept to drive. Then, if the water level is lowered, water injection is started again by the ON-OFF for each of the valves. (Ikeda, J.)

Reactor power measuring device

International Nuclear Information System (INIS)

Izumi, Mikio; Sano, Yuji; Seki, Eiji; Yoshida, Toshifumi; Ito, Toshiaki.

1993-01-01

The present invention provides a self-powered long detector having a sensitivity over the entire length of a reactor core as an entire control rod withdrawal range of a BWR type reactor, and a reactor power measuring device using a gamma ray thermometer which scarcely causes sensitivity degradation. That is, a hollow protection pipe is disposed passing through the reactor core from the outside of a reactor pressure vessel. The self-powered long detectors and the gamma ray thermometers are inserted and installed in the protection pipe. An average reactor power in an axial direction of the reactor relative to a certain position in the horizontal cross section of the reactor core is determined based on the power of the self-powered long detector over the entire length of the reactor core. Since the response of the self-powered detector relative to a local power change is rapid, the output is used as an input signal to a safety protection device of the reactor core. Further, a gamma ray thermometer secured in the reactor and having scarce sensitivity degradation is used instead of an incore travelling neutron monitor used for relative calibration of an existent neutron monitor secured in the reactor. (I.S.)

Status of high-temperature heat-pipe technology

International Nuclear Information System (INIS)

Ranken, W.A.

1982-01-01

This paper discusses the application of heat pipes to nuclear reactor space power systems. Characteristics of the device that favor such an application are described and recent results of current technology development programs are presented. Research areas that will need to be addressed in demonstrating that adequate lifetimes can be achieved with evaporation/condensation cycles operating at high temperatures in a reactor environment are also discussed

Damping considerations in CANDU feeder pipe design and analysis

International Nuclear Information System (INIS)

Usmani, S.A.; Saleem, M.A.; So, G.

1990-01-01

Recent developments in pipe damping indicate a trend towards more realistic and less conservative values, which result in less rigid and safer pipe designs. The CANDU-PHW (Canada deuterium uranium, pressurized heavy water) reactor feeder pipe designs have applied similar approaches which permit seismic qualifications without overly restraining these compact arrays of pipes to cater for the large creep and thermal anchor movement. This paper reviews the feeder design aspects, especially pertaining to the design provisions, experimental verification and analytical modelling for seismic qualification in the light of recent pipe dynamic developments. Using illustrative examples, comparison of seismic analysis results is provided for the ASME Code Case N-411 dampings, and those traditionally used in the feeder seismic qualification. The results confirm acceptability of the traditional approach which permit simplified analysis to demonstrate seismic qualificationqualification of CANDU feeder pipes

49 CFR 192.311 - Repair of plastic pipe.

Science.gov (United States)

2010-10-01

... 49 Transportation 3 2010-10-01 2010-10-01 false Repair of plastic pipe. 192.311 Section 192.311... Lines and Mains § 192.311 Repair of plastic pipe. Each imperfection or damage that would impair the serviceability of plastic pipe must be repaired or removed. [Amdt. 192-93, 68 FR 53900, Sept. 15, 2003] ...

Inelastic analysis of SNR-300 piping

International Nuclear Information System (INIS)

Huebel, H.; Di Luna, L.J.; Moy, G.

1983-01-01

This paper investigates plasticity, creep, and elastic follow-up effects on a full size hot primary piping system of the German fast breeder reactor prototype, the SNR-300. A large model (327 elements, 419 nodes) of straight pipe, special elbow and hanger elements of the general purpose finite element program, MARC-CDC, is used to predict piping behavior for a heat-up, sodium loading-unloading-reloading cycle and other significant operating conditions. Included in this work are many time-dependent solution increments for a 5,000 hour creep period. Creep strains and relaxed stress results, after 5,000 hours, for the complete model are used with uniaxial and biaxial models and results to extrapolate conclusions for a 100,000 hour operating life. (author)

Inelastic analysis of SNR-300 piping

Energy Technology Data Exchange (ETDEWEB)

Huebel, H [INTERATOM, Bergisch Gladbach (Germany); Di Luna, L J; Moy, G [Teledyne Engineering Services, Waltham, MA (United States)

1983-05-01

This paper investigates plasticity, creep, and elastic follow-up effects on a full size hot primary piping system of the German fast breeder reactor prototype, the SNR-300. A large model (327 elements, 419 nodes) of straight pipe, special elbow and hanger elements of the general purpose finite element program, MARC-CDC, is used to predict piping behavior for a heat-up, sodium loading-unloading-reloading cycle and other significant operating conditions. Included in this work are many time-dependent solution increments for a 5,000 hour creep period. Creep strains and relaxed stress results, after 5,000 hours, for the complete model are used with uniaxial and biaxial models and results to extrapolate conclusions for a 100,000 hour operating life. (author)

Elastic-plastic response of a piping system due to simulated double-ended guillotine break events

International Nuclear Information System (INIS)

Kussmaul, K.; Diem, H.; Hunger, H.; Katzenmeier, G.

1987-01-01

From the blowdown experiments performed on the HDR feedwater line with feedwater check valve the conclusion can be drawn that high transient loads of up to plastic strains of 3%, acting on an initially integer piping system, can be sustained without loss of integrity for a low number of load cycles due to the plasticizing capacity of the pipework materials nowadays used in the reactor technology. In the experiments carried out with ferritic piping of ND 400 pressure peaks up to about 31,5 mPA were achieved which resulted in excessive strains of up to 3%. By nonlinear finite element computations (ABAQUS) it was possible to describe the elastic-plastic behaviour of the piping in a good approximation. On account of the safety margins proved in the experiments, potential inaccuracies in theoretical structure analyses are recommended so as to be on the safe side. On the other hand, it appears that designing pipework with reference to elastic stress categories does not adequately take into account the actual reserves of the pipework material

Comparison of elastic and inelastic seismic response of high temperature piping systems

International Nuclear Information System (INIS)

Thomas, F.M.; McCabe, S.L.; Liu, Y.

1994-01-01

A study of high temperature power piping systems is presented. The response of the piping systems is determined when subjected to seismic disturbances. Two piping systems are presented, a main steam line, and a cold reheat line. Each of the piping systems are modeled using the ANSYS computer program and two analyses are performed on each piping system. First, each piping system is subjected to a seismic disturbance and the pipe material is assumed to remain linear and elastic. Next the analysis is repeated for each piping system when the pipe material is modeled as having elastic-plastic behavior. The results of the linear elastic analysis and elastic-plastic analysis are compared for each of the two pipe models. The pipe stresses, strains, and displacements, are compared. These comparisons are made so that the effect of the material yielding can be determined and to access what error is made when a linear analysis is performed on a system that yields

The influence of prefabricated pipe cement coatings and those made during pipe renovation on drinking water quality

Directory of Open Access Journals (Sweden)

Młyńska Anna

2017-01-01

Full Text Available Nowadays, cement coatings are often used as an anticorrosion protection of the internal surfaces of manufactured ductile iron water pipes. The protective cement linings are also commonly used for old water pipe renovation. In both cases, the cement lining is an excellent anticorrosion protection of the pipelines, effectively separating the pipe wall from the flowing water. Moreover, cement linings protect the pipelines not only by a mechanical barrier, but also by a chemical barrier creating a highly alkaline environment in water contact with the metal pipe wall. In addition, cement coatings have an ability for so-called self-regeneration and provide the improvement of hydraulic conditions inside the pipelines. In turn, the differences between the analysed cement coatings mainly depend on the types of cements used and techniques of cement mortar spraying. As was expected, they influence the quality of water having contact with the coating. A comparison of the impact of cement coatings manufactured in factories and sprayed on building sites during the renovation on drinking water quality parameters was performed in the study. The experiments were conducted in laboratory conditions, using the test stands prepared for this purpose. The results include analysis of selected water quality parameters for the samples contacting with cement mortar and collected during the investigation.

Environmentally assisted cracking of light-water reactor materials

International Nuclear Information System (INIS)

Chopra, O.K.; Chung, H.M.; Kassner, T.F.; Shack, W.J.

1996-02-01

Environmentally assisted cracking (EAC) of lightwater reactor (LWR) materials has affected nuclear reactors from the very introduction of the technology. Corrosion problems have afflicted steam generators from the very introduction of pressurized water reactor (PWR) technology. Shippingport, the first commercial PWR operated in the United States, developed leaking cracks in two Type 304 stainless steel (SS) steam generator tubes as early as 1957, after only 150 h of operation. Stress corrosion cracks were observed in the heat-affected zones of welds in austenitic SS piping and associated components in boiling-water reactors (BRWs) as early as 1965. The degradation of steam generator tubing in PWRs and the stress corrosion cracking (SCC) of austenitic SS piping in BWRs have been the most visible and most expensive examples of EAC in LWRs, and the repair and replacement of steam generators and recirculation piping has cost hundreds of millions of dollars. However, other problems associated with the effects of the environment on reactor structures and components am important concerns in operating plants and for extended reactor lifetimes. Cast duplex austenitic-ferritic SSs are used extensively in the nuclear industry to fabricate pump casings and valve bodies for LWRs and primary coolant piping in many PWRs. Embrittlement of the ferrite phase in cast duplex SS may occur after 10 to 20 years at reactor operating temperatures, which could influence the mechanical response and integrity of pressure boundary components during high strain-rate loading (e.g., seismic events). The problem is of most concern in PWRs where slightly higher temperatures are typical and cast SS piping is widely used

Application of LBB to high energy piping systems in operating PWR

Energy Technology Data Exchange (ETDEWEB)

Swamy, S.A.; Bhowmick, D.C. [Westinghouse Nuclear Technology Division, Pittsburgh, PA (United States)

1997-04-01

The amendment to General Design Criterion 4 allows exclusion, from the design basis, of dynamic effects associated with high energy pipe rupture by application of leak-before-break (LBB) technology. This new approach has resulted in substantial financial savings to utilities when applied to the Pressurized Water Reactor (PWR) primary loop piping and auxiliary piping systems made of stainless steel material. To date majority of applications pertain to piping systems in operating plants. Various steps of evaluation associated with the LBB application to an operating plant are described in this paper.

Study on air ingress during an early stage of a primary-pipe rupture accident of a high-temperature gas-cooled reactor

International Nuclear Information System (INIS)

Hishida, M.; Takeda, T.

1991-01-01

A primary-pipe rupture accident is one of the design-based accidents of the HTTR. As the first step of our final goal of predicting the multicomponent gas flow in a reactor during the early stages of the accident, the present paper aims at studying experimentally and analytically, the basic features of air ingress and gas transportation by transient molecular diffusion and the transient natural convection of a two-component gas mixture. The present paper comprises two main parts. The first part deals with analytical and experimental studies on N 2 ingress (corresponding to air ingress) and gas transportation by molecular diffusion and the one-dimensional natural convection of an He-N 2 two-component gas mixture in a reverse-U-shaped tube. Analytical and experimental results are discussed on the N 2 mole fraction change with time after the simulated pipe rupture and on the initation time of the natural circulation of pure N 2 . The second part deals with a preliminary simulation test of air ingress during the early stages of the accident. The test is performed with a very simple model of the reactor. The experimental results are discussed on the change in mole fraction of air with time and on the initiation time of the natural circulation of pure air. (orig.)

Non-stationary flow of hydraulic oil in long pipe

Directory of Open Access Journals (Sweden)

Hružík Lumír

2014-03-01

Full Text Available The paper deals with experimental evaluation and numerical simulation of non-stationary flow of hydraulic oil in a long hydraulic line. Non-stationary flow is caused by a quick closing of valves at the beginning and the end of the pipe. Time dependence of pressure is measured by means of pressure sensors at the beginning and the end of the pipe. A mathematical model of a given circuit is created using Matlab SimHydraulics software. The long line is simulated by means of segmented pipe. The simulation is verified by experiment.

Characterization of radioactive contamination inside pipes with the Pipe Explorer trademark system

International Nuclear Information System (INIS)

Kendrick, D.T.; Cremer, C.D.; Lowry, W.; Cramer, E.

1995-01-01

The U.S. Department of Energy's nuclear facility decommissioning program needs to characterize radiological contamination inside piping systems before the pipe can be recycled, remediated, or disposed. Science and Engineering associates, Inc. under contract with the DOE Morgantown Energy Technology Center has developed and demonstrated the Pipe Explorer trademark system, which uses an inverting membrane to transport various characterization sensors into pipes. The basic process involves inverting (turning inside out) a tubular impermeable membrane under air pressure. A characterization sensor is towed down the interior of the pipe by the membrane. Advantages of this approach include the capability of deploying through constrictions in the pipe, around 90 degrees bends, vertically up and down, and in slippery conditions. Because the detector is transported inside the membrane (which is inexpensive and disposable), it is protected from contamination, which eliminates cross-contamination. Characterization sensors that have been demonstrated with the system thus far include: gamma detectors, beta detectors, video cameras, and pipe locators. Alpha measurement capability is currently under development. A remotely operable Pipe Explorer trademark system has been developed and demonstrated for use in DOE facilities in the decommissioning stage. The system is capable of deployment in pipes as small as 2-inch-diameter and up to 250 feet long. This paper describes the technology and presents measurement results of a field demonstration conducted with the Pipe Explorer trademark system at a DOE site. These measurements identify surface activity levels of U-238 contamination as a function of location in drain lines. Cost savings to the DOE of approximately $1.5 million dollars were realized from this one demonstration

On-line system for monitoring of boiling in nuclear reactor fuel assemblies

International Nuclear Information System (INIS)

Tuerkcan, E.; Kozma, R.; Verhoef, J.P.; Nabeshima, K.

1996-01-01

An important goal of nuclear reactor instrumentation is the continuous monitoring of the state of the reactor and the detection of deviations from the normal behaviour at an early stage. Early detection of anomalies enables one to make the necessary steps in order to prevent further damage of nuclear fuel. In the present paper, an on-line core monitoring system is described by means of which boiling anomaly in nuclear reactor fuel assemblies can be detected. (author). 9 refs, 7 figs

Condensation driven water hammer studies for feedwater distribution pipe

Energy Technology Data Exchange (ETDEWEB)

Savolainen, S.; Katajala, S.; Elsing, B.; Nurkkala, P.; Hoikkanen, J. [Imatran Voima Oy, Vantaa (Finland); Pullinen, J. [IVO Power Engineering Ltd., Vantaa (Finland); Logvinov, S.A.; Trunov, N.B.; Sitnik, J.K. [EDO Gidropress (Russian Federation)

1997-12-31

Imatran Voima Oy, IVO, operates two VVER 440 reactors. Unit 1 has been operating since 1977 and unit 2 since 1981. First damages of the feed water distribution (FWD) pipes were observed in 1989. In closer examinations FWD-pipe T-connection turned out to suffer from severe erosion corrosion damages. Similar damages have been found also in other VVER 440 type NPPs. In 1994 the first new FWD-pipe was replaced and in 1996 extensive water hammer experiments were carried out together with EDO Gidropress in Podolsk. After the first phase of the experiments some fundamental changes were made to the construction of the FWD-pipe. The object of this paper is to give short insight to the design of the new FWD-pipe concentrating on water hammer experiments. (orig.).

Condensation driven water hammer studies for feedwater distribution pipe

International Nuclear Information System (INIS)

Savolainen, S.; Katajala, S.; Elsing, B.; Nurkkala, P.; Hoikkanen, J.; Pullinen, J.; Logvinov, S.A.; Trunov, N.B.; Sitnik, J.K.

1997-01-01

Imatran Voima Oy, IVO, operates two VVER 440 reactors. Unit 1 has been operating since 1977 and unit 2 since 1981. First damages of the feed water distribution (FWD) pipes were observed in 1989. In closer examinations FWD-pipe T-connection turned out to suffer from severe erosion corrosion damages. Similar damages have been found also in other VVER 440 type NPPs. In 1994 the first new FWD-pipe was replaced and in 1996 extensive water hammer experiments were carried out together with EDO Gidropress in Podolsk. After the first phase of the experiments some fundamental changes were made to the construction of the FWD-pipe. The object of this paper is to give short insight to the design of the new FWD-pipe concentrating on water hammer experiments. (orig.)

Condensation driven water hammer studies for feedwater distribution pipe

Energy Technology Data Exchange (ETDEWEB)

Savolainen, S; Katajala, S; Elsing, B; Nurkkala, P; Hoikkanen, J [Imatran Voima Oy, Vantaa (Finland); Pullinen, J [IVO Power Engineering Ltd., Vantaa (Finland); Logvinov, S A; Trunov, N B; Sitnik, J K [EDO Gidropress (Russian Federation)

1998-12-31

Imatran Voima Oy, IVO, operates two VVER 440 reactors. Unit 1 has been operating since 1977 and unit 2 since 1981. First damages of the feed water distribution (FWD) pipes were observed in 1989. In closer examinations FWD-pipe T-connection turned out to suffer from severe erosion corrosion damages. Similar damages have been found also in other VVER 440 type NPPs. In 1994 the first new FWD-pipe was replaced and in 1996 extensive water hammer experiments were carried out together with EDO Gidropress in Podolsk. After the first phase of the experiments some fundamental changes were made to the construction of the FWD-pipe. The object of this paper is to give short insight to the design of the new FWD-pipe concentrating on water hammer experiments. (orig.).

A PC-based high temperature gas reactor simulator for Indonesian conceptual HTR reactor basic training

Science.gov (United States)

Syarip; Po, L. C. C.

2018-05-01

In planning for nuclear power plant construction in Indonesia, helium cooled high temperature reactor (HTR) is favorable for not relying upon water supply that might be interrupted by earthquake. In order to train its personnel, BATAN has cooperated with Micro-Simulation Technology of USA to develop a 200 MWt PC-based simulation model PCTRAN/HTR. It operates in Win10 environment with graphic user interface (GUI). Normal operation of startup, power maneuvering, shutdown and accidents including pipe breaks and complete loss of AC power have been conducted. A sample case of safety analysis simulation to demonstrate the inherent safety features of HTR was done for helium pipe break malfunction scenario. The analysis was done for the variation of primary coolant pipe break i.e. from 0,1% - 0,5 % and 1% - 10 % helium gas leakages, while the reactor was operated at the maximum constant power of 10 MWt. The result shows that the highest temperature of HTR fuel centerline and coolant were 1150 °C and 1296 °C respectively. With 10 kg/s of helium flow in the reactor core, the thermal power will back to the startup position after 1287 s of helium pipe break malfunction.

Development of a portable computed tomographic scanner for on-line imaging of industrial piping systems

International Nuclear Information System (INIS)

Jaafar Abdullah; Mohd Arif Hamzah; Mohd Soyapi Mohd Yusof; Mohd Fitri Abdul Rahman; Fadil IsmaiI; Rasif Mohd Zain

2003-01-01

Computed tomography (CT) technology is being increasingly developed for industrial application. This paper presents the development of a portable computed tomographic scanner for on?line imaging of industrial piping systems. The theoretical approach, the system hardware, the data acquisition system and the adopted algorithm for image reconstruction are discussed. The scanner has large potential to be used to determine the extent of corrosion under insulation (CUI), to detect blockages, to measure the thickness of deposit/materials built-up on the walls and to improve understanding of material flow in pipelines. (Author)

Some scoping experiments for a space reactor

International Nuclear Information System (INIS)

Alexander, C.A.; Ogden, J.S.

1983-01-01

Some scoping experiments were performed to evaluate fuel performance in a lithium heat pipe reactor operating at a nominal 1500K heat pipe temperature. Fuel-coolant and fuel-coolant-clad relationships showed that once a failed heat pipe occurs temperatures can rise high enough so that large concentrations of uranium can be transported by the vapor phase. Upon condensation this uranium would be capable of penetrating heat pipes adjacent to the failed pipe. The potential for propagation of failure exists with UO 2 and a lithium heat pipe. Changing the composition of the metal of the heat pipe would have only a second order effect on the kinetics of the failure mechanism. Uranium carbide and nitride were considered as potential fuels which are nonreactive in a lithium environment. At high temperatures the nitride would be favored because of its better compatibility with potential cladding materials. Compositions of UN with small additions of YN appear to offer very attractive properties for a compact high temperature high power density reactor

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

Compact reactor/ORC power source

International Nuclear Information System (INIS)

Meier, K.L.; Kirchner, W.L.; Willcutt, G.J.

1986-01-01

A compact power source that combines an organic Rankine Cycle (ORC) electric generator with a nuclear reactor heat source is being designed and fabricated. Incorporating existing ORC technology with proven reactor technology, the compact reactor/ORC power source offers high reliability while minimizing the need for component development. Thermal power at 125 kWt is removed from the coated particle fueled, graphite moderated reactor by heat pipes operating at 500 0 C. Outside the reactor vessel and connected to the heat pipes are vaporizers in which the toluene ORC working fluid is heated to 370 0 C. In the turbine-alternator-pump (TAP) combined-rotating unit, the thermal energy of the toluene is converted to 25 kWe of electric power. Lumped parameter systems analyses combined with a finite element thermal analysis have aided in the power source design. The analyses have provided assurance of reliable multiyear normal operation as well as full power operation with upset conditions, such as failed heat pipes and inoperative ORC vaporizers. Because of inherent high reliability, long life, and insensitivity to upset conditions, this power source is especially suited for use in remote, inaccessible locations where fuel delivery and maintenance costs are high. 10 refs

Performance demonstration of a high-power space-reactor heat-pipe design

International Nuclear Information System (INIS)

Merrigan, M.A.; Martinez, E.H.; Keddy, E.S.; Runyan, J.; Kemme, J.E.

1983-01-01

Performance of a 15.9-mm diam, 2-m long, artery heat pipe has been demonstrated at power levels to 22.6 kW and temperatures to 1500 0 K. The heat pipe employed lithium as a working fluid with distribution wicks and arteries fabricated from 400 mesh Mo-41 wt % Re screen. Molybdenum alloy (TZM) was used for the container. Peak axial power density attained in the testing was 19 kW/cm 2 at 1465 0 K. The corresponding radial flux density in the evaporator region of the heat pipe was 150 W/cm 2 . The extrapolated limit for the heat pipe at its 1500 0 K design point is 30 kW, corresponding to an axial flux density of 25 kW/cm 2 . Sonic and capillary limits for the design were investigated in the 1100 to 1500 0 K temperature range. Excellent agreement of measured and predicted temperature and power levels was observed

Experimental investigation on Heat Transfer Performance of Annular Flow Path Heat Pipe

International Nuclear Information System (INIS)

Kim, In Guk; Kim, Kyung Mo; Jeong, Yeong Shin; Bang, In Cheol

2015-01-01

Mochizuki et al. was suggested the passive cooling system to spent nuclear fuel pool. Detail analysis of various heat pipe design cases was studied to determine the heat pipes cooling performance. Wang et al. suggested the concept PRHRS of MSR using sodium heat pipes, and the transient performance of high temperature sodium heat pipe was numerically simulated in the case of MSR accident. The meltdown at the Fukushima Daiichi nuclear power plants alarmed to the dangers of station blackout (SBO) accident. After the SBO accident, passive decay heat removal systems have been investigated to prevent the severe accidents. Mochizuki et al. suggested the heat pipes cooling system using loop heat pipes for decay heat removal cooling and analysis of heat pipe thermal resistance for boiling water reactor (BWR). The decay heat removal systems for pressurized water reactor (PWR) were suggested using natural convection mechanisms and modification of PWR design. Our group suggested the concept of a hybrid heat pipe with control rod as Passive IN-core Cooling System (PINCs) for decay heat removal for advanced nuclear power plant. Hybrid heat pipe is the combination of the heat pipe and control rod. In the present research, the main objective is to investigate the effect of the inner structure to the heat transfer performance of heat pipe containing neutron absorber material, B 4 C. The main objective is to investigate the effect of the inner structure in heat pipe to the heat transfer performance with annular flow path. ABS pellet was used instead of B 4 C pellet as cylindrical structures. The thermal performances of each heat pipes were measured experimentally. Among them, concentric heat pipe showed the best performance compared with others. 1. Annular evaporation section heat pipe and annular flow path heat pipe showed heat transfer degradation. 2. AHP also had annular vapor space and contact cooling surface per unit volume of vapor was increased. Heat transfer coefficient of

Evaluation of the Webler-Brown model for estimating tetrachloroethylene exposure from vinyl-lined asbestos-cement pipes

Directory of Open Access Journals (Sweden)

Heeren Timothy C

2008-06-01

Full Text Available Abstract Background From May 1968 through March 1980, vinyl-lined asbestos-cement (VL/AC water distribution pipes were installed in New England to avoid taste and odor problems associated with asbestos-cement pipes. The vinyl resin was applied to the inner pipe surface in a solution of tetrachloroethylene (perchloroethylene, PCE. Substantial amounts of PCE remained in the liner and subsequently leached into public drinking water supplies. Methods Once aware of the leaching problem and prior to remediation (April-November 1980, Massachusetts regulators collected drinking water samples from VL/AC pipes to determine the extent and severity of the PCE contamination. This study compares newly obtained historical records of PCE concentrations in water samples (n = 88 with concentrations estimated using an exposure model employed in epidemiologic studies on the cancer risk associated with PCE-contaminated drinking water. The exposure model was developed by Webler and Brown to estimate the mass of PCE delivered to subjects' residences. Results The mean and median measured PCE concentrations in the water samples were 66 and 0.5 μg/L, respectively, and the range extended from non-detectable to 2432 μg/L. The model-generated concentration estimates and water sample concentrations were moderately correlated (Spearman rank correlation coefficient = 0.48, p Conclusion PCE concentration estimates generated using the Webler-Brown model were moderately correlated with measured water concentrations. The present analysis suggests that the exposure assessment process used in prior epidemiological studies could be improved with more accurate characterization of water flow. This study illustrates one method of validating an exposure model in an epidemiological study when historical measurements are not available.

Heat dissipating nuclear reactor

Science.gov (United States)

Hunsbedt, A.; Lazarus, J.D.

1985-11-21

Disclosed is a nuclear reactor containment adapted to retain and cool core debris in the unlikely event of a core meltdown and subsequent breach in the reactor vessel. The reactor vessel is seated in a cavity which has a thick metal sidewall that is integral with a thick metal basemat at the bottom of the cavity. The basemat extends beyond the perimeter of the cavity sidewall. Underneath the basemat is a porous bed with water pipes and steam pipes running into it. Water is introduced into the bed and converted into steam which is vented to the atmosphere. A plurality of metal pilings in the form of H-beams extend from the metal base plate downwardly and outwardly into the earth.

75 FR 39680 - Houston Pipe Line Company LP, Worsham-Steed Gas Storage, L.P., Energy Transfer Fuel, LP, Mid...

Science.gov (United States)

2010-07-12

... DEPARTMENT OF ENERGY Federal Energy Regulatory Commission [Docket No. PR10-44-000; Docket No. PR10-46-000; Docket No. PR10-48- 000; Docket No. PR10-49-000; Docket No. PR10-50-000] Houston Pipe Line Company LP, Worsham-Steed Gas Storage, L.P., Energy Transfer Fuel, LP, Mid Continent Market Center, L.L.C...

Analysis of residual stresses in girth welded type 304 stainless steel pipes

International Nuclear Information System (INIS)

Brust, F.W.; Kanninen, M.F.

1981-01-01

Intergranular stress corrosion cracking (IGSCC) in boiling water reactor (BWR) piping is a problem for the nuclear power industry. Tensile residual stresses induced by welding are an important factor in IGSCC of Type 304 stainless steel pipes. Backlay and heat sink welding can retard IGSCC. 17 refs

Program to justify life extension of older nuclear piping systems

International Nuclear Information System (INIS)

Burr, T.K.; Dwight, J.E. Jr.; Morton, D.K.

1991-01-01

The Idaho National Engineering Laboratory (INEL) has a history of more than 40 years devoted to the operation of nuclear reactors designed for research and experiments. The Advanced Test Reactor (ATR) is one such operating reactor whose mission requires continued operation for an additional 25 years or more. Since the ATR is approaching its design life of twenty years, life extension evaluations have been initiated. Of particular importance are the associated high temperature, high pressure loop piping system supporting in--reactor experiments. Failure of this piping could challenge core safety margins. Since regulatory rules for nuclear power plant life extension are only in the formulation stage, the current technical guidance on this subject provided by the Department of Energy (DOE) or the commercial nuclear industry is incomplete. In the interim, order to assure continued safe operation of this piping beyond its initial design life, a program has been developed to provide the necessary technical justification for life extension. This paper describes a program that establishes Section 11 of the ASME Boiler and Pressure Vessel Code as the governing criteria document, retains B31.1 as the Code of record for Section 11 activities, specifies additional inservice inspection requirements more strict than Section 11, and relies heavily on flaw detection and fracture mechanics evaluations. 18 refs., 2 figs

The Canadian approach to protection against postulated primary heat transport piping failures

International Nuclear Information System (INIS)

Jarman, B.L.

1985-10-01

In Canada, the Atomic Energy Control Act and Regulations stipulate in broad terms the requirements to be met by licensees. In addition, AECB staff have prepared licensing guides to amplify those requirements. For nuclear reactors, these include providing suitable protection against the consequences of failure of any pipe in the reactor cooling system. The suggested means of limiting the damage caused by whipping pipes or steam jets is by separation of equipment, installing barriers, or restraining piping. If, however, the designer can demonstrate that restraints are impractical or detrimental to safety, AECB staff may consider alternate arguments based on a demonstration that piping is likely to crack and then leak for a long time prior to rupture. This alternative approach would not be considered for ruptures having a high probability of defeating containment, damaging essential safety systems, or of disrupting flow to the core to the extent that fuel cooling could not be maintained

30 CFR 75.1905-1 - Diesel fuel piping systems.

Science.gov (United States)

2010-07-01

... facility. (g) Diesel fuel piping systems from the surface shall only be used to transport diesel fuel... storage facility. (h) The diesel fuel piping system must not be located in a borehole with electric power... entry as electric cables or power lines. Where it is necessary for piping systems to cross electric...

Performance Evaluation of the Concept of Hybrid Heat Pipe as Passive In-core Cooling Systems for Advanced Nuclear Power Plant

International Nuclear Information System (INIS)

Jeong, Yeong Shin; Kim, Kyung Mo; Kim, In Guk; Bang, In Cheol

2015-01-01

As an arising issue for inherent safety of nuclear power plant, the concept of hybrid heat pipe as passive in-core cooling systems was introduced. Hybrid heat pipe has unique features that it is inserted in core directly to remove decay heat from nuclear fuel without any changes of structures of existing facilities of nuclear power plant, substituting conventional control rod. Hybrid heat pipe consists of metal cladding, working fluid, wick structure, and neutron absorber. Same with working principle of the heat pipe, heat is transported by phase change of working fluid inside metal cask. Figure 1 shows the systematic design of the hybrid heat pipe cooling system. In this study, the concept of a hybrid heat pipe was introduced as a Passive IN-core Cooling Systems (PINCs) and demonstrated for internal design features of heat pipe containing neutron absorber. Using a commercial CFD code, single hybrid heat pipe model was analyzed to evaluate thermal performance in designated operating condition. Also, 1-dimensional reactor transient analysis was done by calculating temperature change of the coolant inside reactor pressure vessel using MATLAB. As a passive decay heat removal device, hybrid heat pipe was suggested with a concept of combination of heat pipe and control rod. Hybrid heat pipe has distinct feature that it can be a unique solution to cool the reactor when depressurization process is impossible so that refueling water cannot be injected into RPV by conventional ECCS. It contains neutron absorber material inside heat pipe, so it can stop the reactor and at the same time, remove decay heat in core. For evaluating the concept of hybrid heat pipe, its thermal performance was analyzed using CFD and one-dimensional transient analysis. From single hybrid heat pipe simulation, the hybrid heat pipe can transport heat from the core inside to outside about 18.20 kW, and total thermal resistance of hybrid heat pipe is 0.015 .deg. C/W. Due to unique features of long heat

On-line analysis of reactor noise using time-series analysis

International Nuclear Information System (INIS)

McGevna, V.G.

1981-10-01

A method to allow use of time series analysis for on-line noise analysis has been developed. On-line analysis of noise in nuclear power reactors has been limited primarily to spectral analysis and related frequency domain techniques. Time series analysis has many distinct advantages over spectral analysis in the automated processing of reactor noise. However, fitting an autoregressive-moving average (ARMA) model to time series data involves non-linear least squares estimation. Unless a high speed, general purpose computer is available, the calculations become too time consuming for on-line applications. To eliminate this problem, a special purpose algorithm was developed for fitting ARMA models. While it is based on a combination of steepest descent and Taylor series linearization, properties of the ARMA model are used so that the auto- and cross-correlation functions can be used to eliminate the need for estimating derivatives. The number of calculations, per iteration varies lineegardless of the mee 0.2% yield strength displayed anisotropy, with axial and circumferential values being greater than radial. For CF8-CPF8 and CF8M-CPF8M castings to meet current ASME Code S acid fuel cells

Survey on application of probabilistic fracture mechanics approach to nuclear piping

International Nuclear Information System (INIS)

Kashima, Koichi

1987-01-01

Probabilistic fracture mechanics (PFM) approach is newly developed as one of the tools to evaluate the structural integrity of nuclear components. This report describes the current status of PFM studies for pressure vessel and piping system in light water reactors and focuses on the investigations of the piping failure probability which have been undertaken by USNRC. USNRC reevaluates the double-ended guillotine break (DEGB) of rector coolant piping as a design basis event for nuclear power plant by using the PFM approach. For PWR piping systems designed by Westinghouse, two causes of pipe break are considered: pipe failure due to the crack growth and pipe failure indirectly caused by failure of component supports due to an earthquake. PFM approach shows that the probability of DEGB from either cause is very low and that the effect of earthquake on pipe failure can be neglected. (author)

Pipe Explorer surveying system. Innovative technology summary report

International Nuclear Information System (INIS)

1999-06-01

The US Department of Energy's (DOE) Chicago Operations Office and the DOE's Federal Energy Technology Center (FETC) developed a Large Scale Demonstration Project (LSDP) at the Chicago Pile-5 Research Reactor (CP-5) at Argonne National Laboratory-East (ANL). The objective of the LSDP is to demonstrate potentially beneficial decontamination and decommissioning (D and D) technologies in comparison with current baseline technologies. The Pipe Explorer trademark system was developed by Science and Engineering Associates, Inc. (SEA), Albuquerque, NM as a deployment method for transporting a variety of survey tools into pipes and ducts. Tools available for use with the system include alpha, beta and gamma radiation detectors; video cameras; and pipe locator beacons. Different versions of this technology have been demonstrated at three other sites; results of these demonstrations are provided in an earlier Innovative Technology Summary Report. As part of a D and D project, characterization radiological contamination inside piping systems is necessary before pipes can be recycled, remediated or disposed. This is usually done manually by surveying over the outside of the piping only, with limited effectiveness and risk of worker exposure. The pipe must be accessible to workers, and embedded pipes in concrete or in the ground would have to be excavated at high cost and risk of exposure to workers. The advantage of the Pipe Explorer is its ability to perform in-situ characterization of pipe internals

Piping reliability improvement through passive seismic supports

International Nuclear Information System (INIS)

Baltus, R.; Rubbers, A.

1999-01-01