Method for temporary shielding of reactor vessel internals

International Nuclear Information System (INIS)

Grimm, N.P.; Sejvar, J.

1991-01-01

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

Nuclear reactor vessel decontamination systems

International Nuclear Information System (INIS)

McGuire, P. J.

1985-01-01

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

Nuclear reactor vessel fuel thermal insulating barrier

Science.gov (United States)

Keegan, C. Patrick; Scobel, James H.; Wright, Richard F.

2013-03-19

The reactor vessel of a nuclear reactor installation which is suspended from the cold leg nozzles in a reactor cavity is provided with a lower thermal insulating barrier spaced from the reactor vessel that has a hemispherical lower section that increases in volume from the center line of the reactor to the outer extent of the diameter of the thermal insulating barrier and smoothly transitions up the side walls of the vessel. The space between the thermal insulating harrier and the reactor vessel forms a chamber which can be flooded with cooling water through passive valving to directly cool the reactor vessel in the event of a severe accident. The passive inlet valve for the cooling water includes a buoyant door that is normally maintained sealed under its own weight and floats open when the cavity is Hooded. Passively opening steam vents are also provided.

Reactor Vessel Surveillance Program for Advanced Reactor

Energy Technology Data Exchange (ETDEWEB)

Jeong, Kyeong-Hoon; Kim, Tae-Wan; Lee, Gyu-Mahn; Kim, Jong-Wook; Park, Keun-Bae; Kim, Keung-Koo

2008-10-15

This report provides the design requirements of an integral type reactor vessel surveillance program for an integral type reactor in accordance with the requirements of Korean MEST (Ministry of Education, Science and Technology Development) Notice 2008-18. This report covers the requirements for the design of surveillance capsule assemblies including their test specimens, test block materials, handling tools, and monitors of the surveillance capsule neutron fluence and temperature. In addition, this report provides design requirements for the program for irradiation surveillance of reactor vessel materials, a layout of specimens and monitors in the surveillance capsule, procedures of installation and retrieval of the surveillance capsule assemblies, and the layout of the surveillance capsule assemblies in the reactor.

Reactor water spontaneous circulation structure in reactor pressure vessel

International Nuclear Information System (INIS)

Takahashi, Kazumi

1998-01-01

The gap between the inner wall of a reactor pressure vessel of a BWR type reactor and a reactor core shroud forms a down comer in which reactor water flows downwardly. A feedwater jacket to which feedwater at low temperature is supplied is disposed at the outer circumference of the pressure vessel just below a gas/water separator. The reactor water at the outer circumferential portion just below the air/water separator is cooled by the feedwater jacket, and the feedwater after cooling is supplied to the feedwater entrance disposed below the feedwater jacket by way of a feedwater introduction line to supply the feedwater to the lower portion of the down comer. This can cool the reactor water in the down comer to increase the reactor water density in the down comer thereby forming strong downward flows and promote the recycling of the reactor water as a whole. With such procedures, the reactor water can be recycled stably only by the difference of the specific gravity of the reactor water without using an internal pump. In addition, the increase of the height of the pressure vessel can be suppressed. (I.N.)

Reactor vessel pressure transient protection for pressurized water reactors

International Nuclear Information System (INIS)

Zech, G.

1978-09-01

During the past few years the NRC has been studying the issue of protection of the reactor pressure vessels at Pressurized Water Reactors (PWRs) from transients when the vessels are at a relatively low temperature. This effort was prompted by concerns related to the safety margins available to vessel damage as a result of such events. Nuclear Reactor Regulation Category A Technical Activity No. A-26 was established to set forth the NRC plan for resolution of the generic aspects of this safety issue. The purpose of the report is to document the completion of this generic technical activity

FEASIBILITY OF RECYCLING PLUTONIUM AND MINOR ACTINIDES IN LIGHT WATER REACTORS USING HYDRIDE FUEL

International Nuclear Information System (INIS)

Greenspan, Ehud; Todreas, Neil; Taiwo, Temitope

2009-01-01

The objective of this DOE NERI program sponsored project was to assess the feasibility of improving the plutonium (Pu) and minor actinide (MA) recycling capabilities of pressurized water reactors (PWRs) by using hydride instead of oxide fuels. There are four general parts to this assessment: (1) Identifying promising hydride fuel assembly designs for recycling Pu and MAs in PWRs; (2) Performing a comprehensive systems analysis that compares the fuel cycle characteristics of Pu and MA recycling in PWRs using the promising hydride fuel assembly designs identified in Part 1 versus using oxide fuel assembly designs; (3) Conducting a safety analysis to assess the likelihood of licensing hydride fuel assembly designs; and (4) Assessing the compatibility of hydride fuel with cladding materials and water under typical PWR operating conditions Hydride fuel was found to offer promising transmutation characteristics and is recommended for further examination as a possible preferred option for recycling plutonium in PWRs

FEASIBILITY OF RECYCLING PLUTONIUM AND MINOR ACTINIDES IN LIGHT WATER REACTORS USING HYDRIDE FUEL

Energy Technology Data Exchange (ETDEWEB)

Greenspan, Ehud; Todreas, Neil; Taiwo, Temitope

2009-03-10

The objective of this DOE NERI program sponsored project was to assess the feasibility of improving the plutonium (Pu) and minor actinide (MA) recycling capabilities of pressurized water reactors (PWRs) by using hydride instead of oxide fuels. There are four general parts to this assessment: 1) Identifying promising hydride fuel assembly designs for recycling Pu and MAs in PWRs 2) Performing a comprehensive systems analysis that compares the fuel cycle characteristics of Pu and MA recycling in PWRs using the promising hydride fuel assembly designs identified in Part 1 versus using oxide fuel assembly designs 3) Conducting a safety analysis to assess the likelihood of licensing hydride fuel assembly designs 4) Assessing the compatibility of hydride fuel with cladding materials and water under typical PWR operating conditions Hydride fuel was found to offer promising transmutation characteristics and is recommended for further examination as a possible preferred option for recycling plutonium in PWRs.

Reactor vessel decommissioning project. Final report

International Nuclear Information System (INIS)

Schoonen, D.H.

1984-09-01

This report describes a reactor vessel decommissioning project; it documents and explains the project objectives, scope, performance results, and sodium removal process. The project was successfully completed in FY-1983, within budget and without significant problems or adverse impact on the environment. Waste generated by the operation included the reactor vessel, drained sodium, and liquid, solid, and gaseous wastes which were significantly less than project estimates. Personnel radiation exposures were minimized, such that the project total was one-half the predicted exposure level. Except for the sodium removed, the material remaining in the reactor vessel is essentially the same as when the vessel arrived for processing

Improvement to reactor vessel

International Nuclear Information System (INIS)

1974-01-01

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

Reactor pressure vessel design

International Nuclear Information System (INIS)

Foehl, J.

1998-01-01

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

The uranium zirconium hydride research reactor and its applications in research and education

International Nuclear Information System (INIS)

Chen Wei; Wang Daohua; Jiang Xinbiao; A Jinyan; Yang Jun; Chen Da

2003-01-01

This paper describes briefly the performance, the configuration and the prospects of extensive applications in science, technology and education of the Uranium Zirconium Hydride research reactor in China. (author)

The uranium zirconium hydride research reactor and its applications in research and education

Energy Technology Data Exchange (ETDEWEB)

Chen Wei; Wang Daohua; Jiang Xinbiao; A Jinyan; Yang Jun; Chen Da [Northwest Institute of Nuclear Technology, Xi' an (China)

2003-03-01

This paper describes briefly the performance, the configuration and the prospects of extensive applications in science, technology and education of the Uranium Zirconium Hydride research reactor in China. (author)

Texture and hydride orientation relationship of Zircaloy-4 fuel clad tube during its fabrication for pressurized heavy water reactors

International Nuclear Information System (INIS)

Vaibhaw, Kumar; Rao, S.V.R.; Jha, S.K.; Saibaba, N.; Jayaraj, R.N.

2008-01-01

Zircaloy-4 material is used for cladding tube in pressurized heavy water reactors (PHWRs) of 220 MWe and 540 MWe capacity in India. These tubes are fabricated by using various combinations of thermo-mechanical processes to achieve desired mechanical and corrosion properties. Cladding tube develops crystallographic texture during its fabrication, which has significant influence on its in-reactor performance. Due to radiolytic decomposition of water Zircaloy-4 picks-up hydrogen. This hydrogen in excess of its maximum solubility in reactor operating condition (∼300 deg. C), precipitates as zirconium hydrides causing embrittlement of cladding tube. Hydride orientation in the radial direction of the tube limits the service life and lowers the fuel burn-up in reactor. The orientation of the hydride primarily depends on texture developed during fabrication. A correlation between hydride orientation (F n ) with the texture in the tube during its fabrication has been developed using a second order polynomial. The present work is aimed at quantification and correlation of texture evolved in Zircaloy-4 cladding tube using Kearn's f-parameter during its fabrication process

Texture and hydride orientation relationship of Zircaloy-4 fuel clad tube during its fabrication for pressurized heavy water reactors

Science.gov (United States)

Vaibhaw, Kumar; Rao, S. V. R.; Jha, S. K.; Saibaba, N.; Jayaraj, R. N.

2008-12-01

Zircaloy-4 material is used for cladding tube in pressurized heavy water reactors (PHWRs) of 220 MWe and 540 MWe capacity in India. These tubes are fabricated by using various combinations of thermo-mechanical processes to achieve desired mechanical and corrosion properties. Cladding tube develops crystallographic texture during its fabrication, which has significant influence on its in-reactor performance. Due to radiolytic decomposition of water Zircaloy-4 picks-up hydrogen. This hydrogen in excess of its maximum solubility in reactor operating condition (˜300 °C), precipitates as zirconium hydrides causing embrittlement of cladding tube. Hydride orientation in the radial direction of the tube limits the service life and lowers the fuel burn-up in reactor. The orientation of the hydride primarily depends on texture developed during fabrication. A correlation between hydride orientation ( F n) with the texture in the tube during its fabrication has been developed using a second order polynomial. The present work is aimed at quantification and correlation of texture evolved in Zircaloy-4 cladding tube using Kearn's f-parameter during its fabrication process.

Reactor-vessel-sectioning demonstration

International Nuclear Information System (INIS)

Lundgren, R.A.

1981-09-01

A technical demonstration was successfully completed of simulated reactor vessel sectioning using the combined techniques of air arc gouging and flame cutting. A 4-ft x 3-ft x 9-in. thick sample was fabricated of A36 carbon steel to simulate a reactor vessel wall. A 1/4-in. layer of stainless steel (SS) was tungsten inert gas (TIG)-welded to the carbon steel. Several techniques were considered to section the simulated reactor vessel; air arc gouging was selected to penetrate the stainless steel, and flame cutting was selected to sever the carbon steel. Three sectioning operations were demonstrated. For all three, the operating parameters were the same; but the position of the sample was varied. For the first cut, the sample was placed in a horizontal position, and it was successfully severed from the SS side. For the second cut, the sample was turned over and cut from the carbon steel side. Cutting from the carbon steel side has the advantages of cost reduction

Nuclear reactor with a suspended vessel

International Nuclear Information System (INIS)

Lemercier, Guy.

1977-01-01

This invention relates to a nuclear reactor with a suspended vessel and applies in particular when this is a fast reactor, the core or active part of the reactor being inside the vessel and immersed under a suitable volume of flowing liquid metal to cool it by extracting the calories released by the nuclear fission in the fuel assemblies forming this core [fr

Examination of parameters affecting overload fracture behavior of flaw-tip hydrides in Zr-2.5Nb pressure tubes in Candu reactors

International Nuclear Information System (INIS)

Cui, J.; Shek, G.K.; Wang, Z.R.

2007-01-01

Service-induced flaws in Zr-2.5Nb alloy pressure tubes in Candu (Canada Deuterium Uranium Reactors) nuclear reactors are susceptible to a crack initiation and growth mechanism known as Delayed Hydride Cracking (DHC), which is a repetitive process that involves hydrogen diffusion, hydride precipitation, growth and fracture of a hydride region at the flaw-tip under a constant load. Crack initiation may also occur under another loading condition when the hydride region is subjected to an overload. An overload occurs when the hydride region at the flaw tip is loaded to a stress higher than that at which this region is formed such as when the reactor experiences a transient pressure higher than the normal operating pressure where the hydride region is formed. Flaw disposition requires justification that the hydride region overload will not fracture the hydride region, and initiate DHC. In this work, monotonically increasing load experiments were performed on unirradiated Zr-2.5Nb pressure tube specimens containing simulated debris frets (V-notch) and bearing pad frets (BPF, U-shape notch) to examine overload fracture behavior of flaw-tip hydrides formed under hydride ratcheting conditions. Hydride cracking in the overload tests was detected by the acoustic emission technique and confirmed by post-test metallurgical examination. Test results indicate that the resistance to overload fracture is affected by a number of parameters including hydride formation stress, flaw shape (V-notch vs. BPF) and flaw radius (0.015 mm vs. 0.1 mm). The notch-tip hydride morphologies were examined by optical microscopy and scanning electron microscopy (SEM) which show that they are affected by the hydride formation conditions, resulting in different overload fracture resistance. Finite element stress analyses were also performed to obtain flaw-tip stress distributions for interpretation of the test results. (authors)

Reactor Vessel and Reactor Vessel Internals Segmentation at Zion Nuclear Power Station - 13230

Energy Technology Data Exchange (ETDEWEB)

Cooke, Conrad; Spann, Holger [Siempelkamp Nuclear Services: 5229 Sunset Blvd., (Suite M), West Columbia, SC, 29169 (United States)

2013-07-01

Zion Nuclear Power Station (ZNPS) is a dual-unit Pressurized Water Reactor (PWR) nuclear power plant located on the Lake Michigan shoreline, in the city of Zion, Illinois approximately 64 km (40 miles) north of Chicago, Illinois and 67 km (42 miles) south of Milwaukee, Wisconsin. Each PWR is of the Westinghouse design and had a generation capacity of 1040 MW. Exelon Corporation operated both reactors with the first unit starting production of power in 1973 and the second unit coming on line in 1974. The operation of both reactors ceased in 1996/1997. In 2010 the Nuclear Regulatory Commission approved the transfer of Exelon Corporation's license to ZionSolutions, the Long Term Stewardship subsidiary of EnergySolutions responsible for the decommissioning of ZNPS. In October 2010, ZionSolutions awarded Siempelkamp Nuclear Services, Inc. (SNS) the contract to plan, segment, remove, and package both reactor vessels and their respective internals. This presentation discusses the tools employed by SNS to remove and segment the Reactor Vessel Internals (RVI) and Reactor Vessels (RV) and conveys the recent progress. SNS's mechanical segmentation tooling includes the C-HORCE (Circumferential Hydraulically Operated Cutting Equipment), BMT (Bolt Milling Tool), FaST (Former Attachment Severing Tool) and the VRS (Volume Reduction Station). Thermal segmentation of the reactor vessels will be accomplished using an Oxygen- Propane cutting system. The tools for internals segmentation were designed by SNS using their experience from other successful reactor and large component decommissioning and demolition (D and D) projects in the US. All of the designs allow for the mechanical segmentation of the internals remotely in the water-filled reactor cavities. The C-HORCE is designed to saw seven circumferential cuts through the Core Barrel and Thermal Shield walls with individual thicknesses up to 100 mm (4 inches). The BMT is designed to remove the bolts that fasten the Baffle

Study on an innovative fast reactor utilizing hydride neutron absorber - Final report of phase I study

International Nuclear Information System (INIS)

Konashi, K.; Iwasaki, T.; Itoh, K.; Hirai, M.; Sato, J.; Kurosaki, K.; Suzuki, A.; Matsumura, Y.; Abe, S.

2010-01-01

These days, the demand to use nuclear resources efficiently is growing for long-term energy supply and also for solving the green house problem. It is indispensable to develop technologies to reduce environmental load with the nuclear energy supply for sustainable development of human beings. In this regard, the development of the fast breeder reactor (FBR) is preferable to utilize nuclear resources effectively and also to burn minor actinides which possess very long toxicity for more than thousands years if they are not extinguished. As one of the FBR developing works in Japan this phase I study started in 2006 to introduce hafnium (Hf) hydride and Gadolinium-Zirconium (Gd-Zr) hydride as new control materials in FBR. By adopting them, the FBR core control technology is improved by two ways. One is extension of control rod life time by using long life Hf hydride which leads to reduce the fabrication and disposal cost and the other is reduction of the excess reactivity by adopting Gd-Zr hydride which leads to reduce the number of control rods and simplifies the core upper structure. This three year study was successfully completed and the following results were obtained. The core design was performed to examine the applicability of the Hf hydride absorber to Japanese Sodium Fast Reactor (JSFR) and it is clarified that the control rod life time can be prolonged to 6 years by adopting Hf hydride and the excess reactivity of the beginning of the core cycle can be reduced to half and the number of the control rods is also reduced to half by using the Gd-Zr hydride burnable poison. The safety analyses also certified that the core safety can be maintained with the same reliability of JSFR Hf hydride and Gd-Zr hydride pellets were fabricated in good manner and their basic features for design use were measured by using the latest devices such as SEM-EDX. In order to reduce the hydrogen transfer through the stainless steel cladding a new technique which shares calorizing

Computerized reactor pressure vessel materials information system

International Nuclear Information System (INIS)

Strosnider, J.; Monserrate, C.; Kenworthy, L.D.; Tether, C.D.

1980-10-01

A computerized information system for storage and retrieval of reactor pressure vessel materials data was established, as part of Task Action Plan A-11, Reactor Vessel Materials Toughness. Data stored in the system are necessary for evaluating the resistance of reactor pressure vessels to flaw-induced fracture. This report includes (1) a description of the information system; (2) guidance on accessing the system; and (3) a user's manual for the system

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

International Nuclear Information System (INIS)

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

2008-01-01

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

Some aspects of reactor pressure vessel integrity

International Nuclear Information System (INIS)

Korosec, D.; Vojvodic, G.J.

1996-01-01

Reactor pressure vessel of the pressurized water reactor nuclear power plant is the subject of extreme interest due to the fact that presents the pressure boundary of the reactor coolant system, which is under extreme thermal, mechanical and irradiation effects. Reactor pressure vessel by itself prevents the release of fission products to the environment. Design, construction and in-service inspection of such component is governed by strict ASME rules and other forms of administrative control. The reactor pressure vessel in nuclear power plant Kriko is designed and constructed in accordance with related ASME rules. The in-service inspection program includes all requests presented in ASME Code section XI. In the present article all major requests for the periodic inspections of reactor pressure vessel and fracture mechanics analysis are discussed. Detailed and strict fulfillment of all prescribed provisions guarantee the appropriate level of nuclear safety. (author)

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

International Nuclear Information System (INIS)

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

2016-01-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

2016-10-15

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

Pressure vessel for nuclear reactor plant consisting of several pre-stressed cast pressure vessels

International Nuclear Information System (INIS)

Bodmann, E.

1984-01-01

Several cylindrical pressure vessel components made of pressure castings are arranged on a sector of a circle around the cylindrical cast pressure vessel for accommodating the helium cooled HTR. Each component pressure vessel is connected to the reactor vessel by a horizontal gas duct. The contact surfaces between reactor and component pressure vessel are in one plane. In the spaces between the individual component pressure vessels, there are supporting blocks made of cast iron, which are hollow and also have flat surfaces. With the reactor vessel and the component pressure vessels they form a disc-shaped connecting part below and above the gas ducts. (orig./PW)

Irradiation embrittlement of reactor vessel steels

International Nuclear Information System (INIS)

Bros, J.

2000-01-01

From the historical decision of closing the Yankee Rowe NPP because of the uncertainties on the level of reactor pressure vessel neutron embrittlement, this paper reviews the technical-scientist bases of the degradation phenomena, and refers to the evolution of reactor pressure vessel radiation surveillance programs. (Author)

Simulation of In-Vessel Corium Retention through External Reactor Vessel Cooling for SMART using SIMPLE

Energy Technology Data Exchange (ETDEWEB)

Jang, Jin-Sung; Son, Donggun; Park, Rae-Joon [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2016-10-15

Thermal load analysis from the corium pool to the outer reactor vessel in the lower plenum of the reactor vessel is necessary to evaluate the effect of the IVR-ERVC during a severe accident for SMART. A computational code called SIMPLE (Sever Invessel Melt Progression in Lower plenum Environment) has been developed for analyze transient behavior of molten corium in the lower plenum, interaction between corium and coolant, and heat-up and ablation of reactor vessel wall. In this study, heat load analysis of the reactor vessel for SMART has been conducted using the SIMPLE. Transient behavior of the molten corium in the lower plenum and IVR-ERVC for SMART has been simulated using SIMPLE. Heat flux from the corium pool to the outer reactor vessel is concentrated in metallic layer by the focusing effect. As a result, metallic layer shows higher temperature than the oxidic layer. Also, vessel wall of metallic layer has been ablated by the high in-vessel temperature. Ex-vessel temperature of the metallic layer was maintained 390 K and vessel thickness was maintained 14 cm. It means that the reactor vessel integrity is maintained by the IVR-ERVC.

Reactor vessel stud closure system

International Nuclear Information System (INIS)

Spiegelman, S.R.; Salton, R.B.; Beer, R.W.; Malandra, L.J.; Cognevich, M.L.

1982-01-01

A quick-acting stud tensioner apparatus for enabling the loosening or tightening of a stud nut on a reactor vessel stud. The apparatus is adapted to engage the vessel stud by closing a gripper around an upper end of the vessel stud when the apparatus is seated on the stud. Upon lifting the apparatus, the gripper releases the vessel stud so that the apparatus can be removed

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

International Nuclear Information System (INIS)

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

2016-01-01

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

Permeation rates for RTF metal hydride vessels

International Nuclear Information System (INIS)

Klein, J.E.

1992-01-01

Contamination rates have been estimated for the RTF nitrogen heating and cooling system (NH and CS) due to tritium permeation through the walls of metal hydride vessels. Tritium contamination of the NH and CS will be seen shortly after start-up of the RTF with the majority of it coming from the TCAP units. Contamination rates of the NH and CS are estimated to exceed 400 Ci/year after three years of operation and will elevate tritium concentrations in the NH and CS above 6 x 10 -3 μCi/cc. To reduce tritium activity in the NH and CS, a stripper or ''getter'' bed may need to be installed in the NH and CS. Increasing the purge rate of nitrogen from the NH and CS is shown to be an impractical method for reducing tritium activity due to the high purge rates required. Stripping of the NH and CS nitrogen in the glove box stripper system will give a temporary lowering of tritium activity in the NH and CS, but tritium activity will return to its previous level in approximately two weeks

Containment vessel construction for nuclear power reactors

International Nuclear Information System (INIS)

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

1975-01-01

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

Results of reactor pressure vessels ISI

International Nuclear Information System (INIS)

Cepcek, S.

1994-01-01

To find out the possible influence of the annealing process to reactor pressure vessel integrity, a large in-service inspection programme has been implemented as an associated activity to reactor pressure vessel annealing. In this paper the approach to the RPV in-service inspection is shown. Also, the main results and conclusions following in-service inspection are presented. (author). 3 refs, 1 fig

Investigation of flow stabilization in a compact reactor vessel of a FBR. Flow visualization in a reactor vessel

International Nuclear Information System (INIS)

Sato, Hiroyuki; Igarashi, Minoru; Kimura, Nobuyuki; Kamide, Hideki

2002-01-01

In the feasibility studies of Commercialized Fast Breeder Reactor Cycle System, a compact reactor vessel is considered from economical improvement point of a sodium cooled loop type fast reactor. The flow field was visualized by water experiment for a reactor vessel with 'a column type UIS (Upper Internal Structure)', which has a slit for fuel handling mechanism and is useful for a compact fast reactor. In this research, the 1/20 scale test equipment using water was made to understand coolant flow through a slit of a column type UIS' and fundamental behavior of reactor upper plenum flow. In the flow visualization tests, tracer particles were added in the water, and illuminated by the slit-shaped pulse laser. The flow visualization image was taken with a CCD camera. We obtained fluid velocity vectors from the visualization image using the Particle Imaging Velocimetry (PIV). The results are as follows. 1. Most of coolant flow through a slit of 'column type UIS' arrived the dip plate directly. In the opposite side of a slit, most of coolant flowed toward reactor vessel wall before it arrived the dip plate. 2. The PIV was useful to measure the flow field in the reactor vessel. The obtained velocity field was consistent with the flow visualization result. 3. The jet through the UIS slit was dependent on the UIS geometry. There is a possibility to control the jet by the UIS geometry. (author)

A study of reactor vessel integrity assessment

Energy Technology Data Exchange (ETDEWEB)

Lee, Jae Hoon [Korea Institute of Nuclear Safety, Taejon (Korea, Republic of); Kim, Jong Kyung; Shin, Chang Ho; Seo, Bo Kyun [Hanyang Univ., Seoul (Korea, Republic of)

1999-02-15

The fast neutron fluence at the Reactor Pressure Vessel(RPV) of KNGR designed for 60 years lifetime was calculated by full-scope Monte Carlo simulation for reactor vessel integrity assessment. KNGR core geometry was modeled on a three-dimensional representation of the one-sixteenth of the reactor in-vessel component. Each fuel assemblies were modeled explicitly, and each fuel pins were axially divided into 5 segments. The maximum flux of 4.3 x 10{sup 10} neutrons/cm{sup 2}. sec at the RPV was obtained by tallying neutrons crossing the beltline of inner surface of the RPV.

Reactor vessel supported by flexure member

International Nuclear Information System (INIS)

Crawford, J.D.; Pankow, B.

1977-01-01

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

Radiation embrittlement of Spanish nuclear reactor pressure vessel steels

International Nuclear Information System (INIS)

Bros, J.; Ballesteros, A.; Lopez, A.

1993-01-01

Commercial pressurized water reactor (PWR) and boiling water reactor (BWR) nuclear power plants contain a series of pressure vessel steel surveillance capsules as the principal means of monitoring radiation effects on the pressure vessel. Changes in fracture toughness are more severe in surveillance capsules than in reactor vessel materials because of their proximity of the reactor core. Therefore, it is possible to predict changes in fracture toughness of the reactor vessel materials. This paper describes the status of the reactor vessel surveillance program relating to Spanish nuclear power plants. To date, twelve capsules have been removed and analyzed from seven of the nine Spanish reactors in operation. The results obtained from the analysis of these capsules are compared with the predictions of the Nuclear Regulatory Commission (NRC) Regulatory Guide 1.99, Rev. 2, by means of measured and expected increase of the nil-ductility transition reference temperature (RT NDT ). The comparison is made considering the different variables normally included in the studies of radiation response of reactor pressure vessel materials, such as copper content of steel, level of neutron fluence above 1 MeV, base metal or weld metal, and so forth. The surveillance data have been used for determining the adjusted reference temperatures and upper shelf energies at any time. The results have shown that the seven pressure vessels are in excellent condition to continue operating with safety against brittle fracture beyond the design life, without the need to recuperate the degraded properties of the materials by annealing of the vessel

Certifying the decommissioned Shippingport reactor vessel for transport

International Nuclear Information System (INIS)

Towell, R.H.

1990-01-01

The decommissioned Shippingport reactor pressure vessel with its concentric neutron shield tank was shipped to Hanford, WA as part of the effort to restore the Shippingport Station to its original condition. The metal walls of the reactor vessel had become radioactive from neutron bombardment while the reactor was operating so it had to be shipped under the regulations for transporting radioactive material. Because of the large amount of radioactivity in the walls, 16,467 Curies, and because the potentially dispersible corrosion layer on the inner walls of both tanks was also radioactive, the Shippingport reactor vessel was transported under the most stringent of the regulations, those for a type B package. Compliance with the packaging regulations was confirmed via independent analysis by the staff of the Department of Energy certifying official and the Shippingport reactor vessel was shipped under DOE Certificate of Compliance USA/9515/B(U)

Dismantling method for reactor pressure vessel and system therefor

International Nuclear Information System (INIS)

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

1994-01-01

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

gamma-Zr-Hydride Precipitate in Irradiated Massive delta- Zr-Hydride

DEFF Research Database (Denmark)

Warren, M. R.; Bhattacharya, D. K.

1975-01-01

During examination of A Zircaloy-2-clad fuel pin, which had been part of a test fuel assembly in a boiling water reactor, several regions of severe internal hydriding were noticed in the upper-plenum end of the pin. Examination of similar fuel pins has shown that hydride of this type is caused by...... to irradiation-induced swelling....

Reactor pressure vessel support

International Nuclear Information System (INIS)

Butti, J.P.

1977-01-01

A link and pin support system provides the primary vertical and lateral support for a nuclear reactor pressure vessel without restricting thermally induced radial and vertical expansion and contraction. (Auth.)

Aging of reactor vessels in LWR type reactors

International Nuclear Information System (INIS)

Gomez Briceno, D.; Lapena, J.; Serrano, M.

2004-01-01

Most of the degradation mechanisms of nuclear components were not included on the design so they have to be treated a posteriori, and that imply a loss of capacity. In this paper the state of the art on the reactor pressure vessel neutron embrittlement and on the irradiation assisted stress corrosion cracking that affects internal components, are explained. Special attention is devoted on the influence of the neutron fluence on IASCC process, on the material alterations promoted by irradiation and their consequences on the susceptibility to this phenomenon. Regarding the reactor pressure vessel degradation, this paper discuss the application of the Master Curve on the structural integrity evaluation of the vessel. Other aspects related to further developments are also mentioned and the importance of a good materials ageing management on the operation of the plant is pointed out. (Author) 12 refs

Prestressed reactor vessel for nuclear power plants

International Nuclear Information System (INIS)

Schoening, J.; Schwiers, H.G.

1982-01-01

With usual pressure vessels for nuclear reactor plants, especially for gas-cooled nuclear reactors, the load occurring due to the inner overpressure, especially the tensile load affecting the vessel top and/or bottom, their axis of inertia being horizontal, shall be compensated without a supplementary modification in design of the top and/or the bottom. This is attained by choosing an appropriate prestressing system of the vessel wall in the field the top and/or the bottom, so that the top and/or the bottom form a tension vault directed towards the interior of the vessel. (orig.) [de

Integrity of pressurized water electronuclear reactor vessels. The case of French reactors

International Nuclear Information System (INIS)

2012-01-01

This document aims at identifying elements related to design, manufacturing and control during operation of reactor vessels of the French electronuclear fleet, and more precisely as far as vessel ferrule is concerned. It briefly describes the typical design and elements of most of French PWR vessels with respect to the reactor type (900 MWe, 1300 MWe, 1450 MWe, EPR). It recalls some measures regarding design (for embrittlement assessment) and manufacturing processes (forging operations for shell fabrication, coatings). It discusses the different manufacturing defects which have been noticed (under the coatings, due to hydrogen, and intergranular loss of cohesion due to re-heating). It more particularly comments defects noticed on a Belgium power station reactor in Doel, defects due to hydrogen and some other defects noticed in the French reactor fleet. It presents the different types of control which are performed on vessel shells during operation

Reactor vessel supported by flexure member

International Nuclear Information System (INIS)

Crawford, J.D.; Pankow, B.

1975-01-01

A description is given of a reactor pressure vessel which is provided with vertical support means in the form of circumferentially spaced columns upon which the vessel is mounted. The columns are adapted to undergo flexure in order to accommodate the thermally induced displacements experienced by the vessel during operational transients

Manipulator arm for a nuclear reactor vessel inspection device

International Nuclear Information System (INIS)

1980-01-01

A manipulator arm for a reactor vessel in-service inspection apparatus is adapted to transport a transducer array for ultrasonic examination of welds at any point in the vessel. The removal of the inspection device from the reactor vessel in an emergency presents a problem where a relatively long manipulator arm is used. This invention provides an improved arm with means for changing the normal orientation of the arm to a shorter one to permit safe removal of the inspection device from the reactor vessel. (author)

Ultrasound periodic inspections of reactor pressure vessels

International Nuclear Information System (INIS)

Haniger, L.

1980-01-01

Two versions are described of ultrasonic equipment for periodic inspections of reactor pressure vessels. One uses the principle of exchangeable programmators with solid-state logic while the other uses programmable logic with semiconductor memories. The equipment is to be used for inspections of welded joints on the upper part of the V-1 reactor pressure vessel. (L.O.)

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

International Nuclear Information System (INIS)

1983-01-01

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

Reactor vessel dismantling at the high flux materials testing reactor Petten

International Nuclear Information System (INIS)

Tas, A.; Teunissen, G.

1986-01-01

The project of replacing the reactor vessel of the high flux materials testing reactor (HFR) originated in 1974 when results of several research programs confirmed severe neutron embrittlement of aluminium alloys suggesting a limited life of the existing facility. This report describes the dismantling philosophy and organisation, the design of special underwater equipment, the dismantling of the reactor vessel and thermal column, and the conditioning and shielding activities resulting in a working area for the installation of the new vessel with no access limitations due to radiation. Finally an overview of the segmentation, waste disposal and radiation exposure is given. The total dismantling, segmentation and conditioning activities resulted in a total collective radiation dose of 300 mSv. (orig.) [de

A water inner circulation device for a reactor vessel

International Nuclear Information System (INIS)

Eriksson, O.

1976-01-01

A water inner circulation device for a reactor vessel comprising a pump mounted in the reactor vessel and driven by a water-cooled electric motor mounted in a housing outside the reactor vessel, the shaft of the pump passing through the reactor-vessel bottom and being coupled to the motor shaft in a member mechanically connected to the bottom of the reactor vessel in the vicinity of the motor housing, the pump shaft being surrounded by a resilient sealing ring, the reactor vessel communicating with the cooling channels of the pump, when the latter is operating, via a slot surrounding the pump hollow cylindrical shaft, characterized in that the slot inner end is used for/forming a circular space surrounding the pump shaft and surrounded by the motorhousing, in which is coaxially mounted a separating cylindral wall, the upper edge of which is tightly applied against the inner wall of the motor-housing to which it is fastened vertically, the inner surface of said wall being turned towards the outer surface of a circular packing-box, the outer surface of said separating wall constituting a separating radical inner surface for a circular chamber through which flow the motor cooling water. (author)

Thermal annealing of an embrittled reactor pressure vessel

International Nuclear Information System (INIS)

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

1998-01-01

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

Integrated conjugate heat transfer analysis method for in-vessel retention with external reactor vessel cooling - 15477

International Nuclear Information System (INIS)

Park, J.W.; Bae, J.H.; Seol, W.C.

2015-01-01

An integrated conjugate heat transfer analysis method for the thermal integrity of a reactor vessel under external reactor vessel cooling conditions is developed to resolve light metal layer focusing effect issue. The method calculates steady-state 3-dimensional temperature distribution of a reactor vessel using coupled conjugate heat transfer between in-vessel 3-layered stratified corium (metallic pool, oxide pool and heavy metal) and polar-angle dependent boiling heat transfer at the outer surface of a reactor vessel. The 3-layer corium heat transfer model is utilizing lumped-parameter thermal-resistance circuit method and ex-vessel boiling regimes are parametrically considered. The thermal integrity of a reactor vessel is addressed in terms of un-molten thickness profile. The vessel 3-dimensional heat conduction is validated against a commercial code. It is found that even though the internal heat flux from the metal layer goes far beyond critical heat flux (CHF) the heat flux from the outermost nodes of the vessel may be maintained below CHF due to massive vessel heat diffusion. The heat diffusion throughout the vessel is more pronounced for relatively low heat generation rate in an oxide pool. Parametric calculations are performed considering thermal conditions such as peak heat flux from a light metal layer, heat generation in an oxide pool and external boiling conditions. The major finding is that the most crucial factor for success of in-vessel retention is not the mass of the molten light metal above the oxide pool but the heat generation rate inside the oxide pool and the 3-dimensional vessel heat transfer provides a much larger minimum vessel wall thickness. (authors)

Hydride embrittlement in zircaloy components

Energy Technology Data Exchange (ETDEWEB)

Lobo, Raquel M.; Andrade, Arnaldo H.P.; Castagnet, Mariano, E-mail: rmlobo@ipen.br [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil)

2011-07-01

Zirconium alloys are used in nuclear reactor cores under high-temperature water environment. During service, hydrogen is generated by corrosion processes, and it is readily absorbed by these materials. When hydrogen concentration exceeds the terminal solid solubility, the excess hydrogen precipitates as zirconium hydride (ZrH{sub 2}) platelets or needles. Zirconium alloys components can fail by hydride cracking if they contain large flaws and are highly stressed. Zirconium alloys are susceptible to a mechanism for crack initiation and propagation termed delayed hydride cracking (DHC). The presence of brittle hydrides, with a K{sub Ic} fracture toughness of only a few MPa{radical}m, results in a severe loss in ductility and toughness when platelet normal is oriented parallel to the applied stress. In plate or tubing, hydrides tend to form perpendicular to the thickness direction due to the texture developed during fabrication. Hydrides in this orientation do not generally cause structural problems because applied stresses in the through-thickness direction are very low. However, the high mobility of hydrogen in a zirconium lattice enables redistribution of hydrides normal to the applied stress direction, which can result in localized embrittlement. When a platelet reaches a critical length it ruptures. If the tensile stress is sufficiently great, crack initiation starts at some of these hydrides. Crack propagation occurs by repeating the same process at the crack tip. Delayed hydride cracking can degrade the structural integrity of zirconium alloys during reactor service. The paper focuses on the fracture mechanics and fractographic aspects of hydride material. (author)

How to replace a reactor pressure vessel

International Nuclear Information System (INIS)

Huber, R.

1996-01-01

A potential life extending procedure for a nuclear reactor after, say, 40 years of service life, might in some circumstances be the replacement of the reactor pressure vessel. Neutron induced degradation of the vessel might make replacement by one of a different material composition desirable, for example. Although the replacement of heavy components, such as steam generators, has been possible for many years, the pressure vessel presents a much more demanding task if only because it is highly irradiated. Some preliminary feasibility studies by Siemens are reported for the two removal strategies that might be considered. These are removal of the entire pressure vessel in one piece and dismantling it into sections. (UK)

Reactor vessel sealing plug

International Nuclear Information System (INIS)

Dooley, R.A.

1986-01-01

An apparatus is described for sealing a cold leg nozzle of a nuclear reactor pressure vessel from a remote location comprising: at least one sealing plug for mechanically sealing the nozzle from the inside of the reactor pressure vessel. The sealing plug includes a plate and a cone assembly having an end part receptive in the nozzle, the plate being axially moveable relative to the cone assembly. The plate and cone assembly have confronting bevelled edges defining an opening therebetween. A primary O-ring is disposed about the opening and is supported on the bevelled edges, the plate being guidably mounted to the cone assembly for movement toward the cone assembly to radially expand the primary O-ring into sealing engagement with the nozzle. A means is included for providing relative movement between the outer plate and the cone assembly

Nuclear reactor pressure vessel flaw distribution development

International Nuclear Information System (INIS)

Kennedy, E.L.; Foulds, J.R.; Basin, S.L.

1991-12-01

Previous attempts to develop flaw distributions for probabilistic fracture mechanics analyses of pressurized water reactor (PWR) vessels have aimed at the estimation of a ''generic'' distribution applicable to all PWR vessels. In contrast, this report describes (1) a new flaw distribution development analytic methodology that can be applied to the analysis of vessel-specific inservice inspection (ISI) data, and (2) results of the application of the methodology to the analysis of flaw data for each vessel case (ISI data on three PWR vessels and laboratory inspection data on sections of the Midland reactor vessel). Results of this study show significant variation among the flaw distributions derived from the various data sets analyzed, strongly suggesting than a vessel-specific flaw distribution (for vessel integrity prediction under pressurized thermal shock) is preferred over a ''generic'' distribution. In addition, quantitative inspection system flaw sizing accuracy requirements have been identified for developing a flaw distribution from vessel ISI data. The new flaw data analysis methodology also permits quantifying the reliability of the flaw distribution estimate. Included in the report are identified needs for further development of several aspects of ISI data acquisition and vessel integrity prediction practice

Fast, quantitative, and nondestructive evaluation of hydrided LWR fuel cladding by small angle incoherent neutron scattering of hydrogen

Energy Technology Data Exchange (ETDEWEB)

Yan, Y.; Qian, S.; Littrell, K.; Parish, C.M. [Oak Ridge National Laboratory, Oak Ridge, TN 37831 (United States); Plummer, L.K. [University of Oregon, Eugene, OR 97403 (United States)

2015-05-15

A nondestructive neutron scattering method to precisely measure the uptake of hydrogen and the distribution of hydride precipitates in light water reactor (LWR) fuel cladding was developed. Zircaloy-4 cladding used in commercial LWRs was used to produce hydrided specimens. The hydriding apparatus consists of a closed stainless-steel vessel that contains Zr alloy specimens and hydrogen gas. Following hydrogen charging, the hydrogen content of the hydrided specimens was measured using the vacuum hot extraction method, by which the samples with desired hydrogen concentrations were selected for the neutron study. Optical microscopy shows that our hydriding procedure results in uniform distribution of circumferential hydrides across the wall thickness. Small angle neutron incoherent scattering was performed in the High Flux Isotope Reactor at Oak Ridge National Laboratory. Our study demonstrates that the hydrogen in commercial Zircaloy-4 cladding can be measured very accurately in minutes by this nondestructive method over a wide range of hydrogen concentrations from a very small amount (≈20 ppm) to over 1000 ppm. The hydrogen distribution in a tube sample was obtained by scaling the neutron scattering rate with a factor determined by a calibration process using standard, destructive direct chemical analysis methods on the specimens. This scale factor can be used in future tests with unknown hydrogen concentrations, thus providing a nondestructive method for determining absolute hydrogen concentrations.

Examination of VVER-1000 Reactor Pressure Vessel

International Nuclear Information System (INIS)

Matokovic, A.; Picek, E.; Markulin, K.

2008-01-01

The increasing demand of a higher level of safety in the operation of the nuclear power plants requires the utilisation of more precise automated equipment to perform in-service inspections. That has been achieved by technological advances in computer technology, in robotics, in examination probe technology with the development of the advanced inspection technique and has also been due to the considerable and varied experience gained in the performance of such inspections. In-service inspection of reactor pressure vessel, especially Russian-designed WWER-1000 presents one of the most important and extensive examination of nuclear power plants primary circuit components. Such examination demand high standards of inspection technology, quality and continual innovation in the field of non-destructive testing advanced technology. A remote underwater contact ultrasonic technique is employed for the examination of the base metal of vessel and reactor welds, whence eddy current method is applied for clad surface examinations. Visual testing is used for examination of the vessel interior. The movement of inspection probes and data positioning are assured by using new reactor pressure vessel tool concept that is fully integrated with inspection systems. The successful performance of reactor pressure vessel is attributed thorough pre-outage planning, training and successful performance demonstration qualification of chosen non-destructive techniques on the specimens with artificial and/or real defects. Furthermore, use of advanced approach of inspection through implementation the state-of-the-art examination equipment significantly reduced the inspection time, radiation exposure to examination personnel, shortening nuclear power plant outage and cutting the total inspection costs. This paper presents advanced approach in the reactor pressure vessel in-service inspections and it is especially developed for WWER-1000 nuclear power plants.(author)

Reactor Pressure Vessel (RPV) Acquisition Strategy

Energy Technology Data Exchange (ETDEWEB)

Mizia, Ronald Eugene [Idaho National Lab. (INL), Idaho Falls, ID (United States)

2008-04-01

The Department of Energy has selected the High Temperature Gas-cooled Reactor design for the Next Generation Nuclear Plant (NGNP) Project. The NGNP will demonstrate the use of nuclear power for electricity and hydrogen production. It will have an outlet gas temperature in the range of 900°C and a plant design service life of 60 years. The reactor design will be a graphite moderated, helium-cooled, prismatic or pebble-bed reactor and use low-enriched uranium, TRISO-coated fuel. The plant size, reactor thermal power, and core configuration will ensure passive decay heat removal without fuel damage or radioactive material releases during accidents. The NGNP Materials Research and Development (R&D) Program is responsible for performing R&D on likely NGNP materials in support of the NGNP design, licensing, and construction activities. Selection of the technology and design configuration for the NGNP must consider both the cost and risk profiles to ensure that the demonstration plant establishes a sound foundation for future commercial deployments. The NGNP challenge is to achieve a significant advancement in nuclear technology while at the same time setting the stage for an economically viable deployment of the new technology in the commercial sector soon after 2020. The purpose of this report is to address the acquisition strategy for the NGNP Reactor Pressure Vessel (RPV). This component will be larger than any nuclear reactor pressure vessel presently in service in the United States. The RPV will be taller, larger in diameter, thicker walled, heavier and most likely fabricated at the Idaho National Laboratory (INL) site of multiple subcomponent pieces. The pressure vessel steel can either be a conventional materials already used in the nuclear industry such as listed within ASME A508/A533 specifications or it will be fabricated from newer pressure vessel materials never before used for a nuclear reactor in the US. Each of these characteristics will present a

Reactor Structural Materials: Reactor Pressure Vessel Steels

International Nuclear Information System (INIS)

Chaouadi, R.

2000-01-01

The objectives of SCK-CEN's R and D programme on Rector Pressure Vessel (RPV) Steels are:(1) to complete the fracture toughness data bank of various reactor pressure vessel steels by using precracked Charpy specimens that were tested statically as well as dynamically; (2) to implement the enhanced surveillance approach in a user-friendly software; (3) to improve the existing reconstitution technology by reducing the input energy (short cycle welding) and modifying the stud geometry. Progress and achievements in 1999 are reported

Reactor vessel assessment and the development of a reactor vessel life extension program for Calvert Cliffs Units One and Two

International Nuclear Information System (INIS)

Montgomery, B.; Hijeck, P.J.

1988-01-01

A study has been undertaken to provide a general assessment of the life extension capabilities for the Calvert Cliffs Units One and Two reactor pressure vessels. The purpose of the study is to assess the general life extension capabilities for the Calvert Cliffs reactor pressure vessels based upon an extension and variation of the Surry pilot plant life extension study. This assessment provided a detailed reactor vessel surveillance program for plant life extension along with a hierarchy of specific tasks necessary for attaining maximum useful life. The assessment identified a number of critical issues which may impact life attainment and extension along with potential solutions to address these issues to ensure the life extension option is not precluded

Application of material databases for improved reliability of reactor pressure vessels

International Nuclear Information System (INIS)

Griesbach, T.J.; Server, W.L.; Beaudoin, B.F.; Burgos, B.N.

1994-01-01

A vital part of reactor vessel Life Cycle Management program must begin with an accurate characterization of the vessel material properties. Uncertainties in vessel material properties or use of bounding values may result in unnecessary conservatisms in vessel integrity calculations. These conservatisms may be eliminated through a better understanding of the material properties in reactor vessels, both in the unirradiated and irradiated conditions. Reactor vessel material databases are available for quantifying the chemistry and Charpy shift behavior of individual heats of reactor vessel materials. Application of the databases for vessels with embrittlement concerns has proven to be an effective embrittlement management tool. This paper presents details of database development and applications which demonstrate the value of using material databases for improving material chemistry and for maximizing the data from integrated material surveillance programs

Prestressed concrete pressure vessels for nuclear reactors - 1973

Energy Technology Data Exchange (ETDEWEB)

1977-01-01

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

The reactor vessel steels

International Nuclear Information System (INIS)

Bilous, W.; Hajewska, E.; Szteke, W.; Przyborska, M.; Wasiak, J.; Wieczorkowski, M.

2005-01-01

In the paper the fundamental steels using in the construction of pressure vessel water reactor are discussed. The properties of these steels as well as the influence of neutron irradiation on its degradation in the time of exploitation are also done. (authors)

Influence of hydrides orientation on strain, damage and failure of hydrided zircaloy-4

International Nuclear Information System (INIS)

Racine, A.

2005-09-01

In pressurized water reactors of nuclear power plants, fuel pellets are contained in cladding tubes, made of Zirconium alloy, for instance Zircaloy-4. During their life in the primary water of the reactor (155 bars, 300 C), cladding tubes are oxidized and consequently hydrided. A part of the hydrogen given off precipitates as Zirconium hydrides in the bulk material and embrittles the material. This embrittlement depends on many parameters, among which hydrogen content and orientation of hydrides with respect to the applied stress. This investigation is devoted to the influence of the orientation of hydrides with respect to the applied stress on strain, damage and failure mechanisms. Macroscopic and SEM in-situ ring tensile tests are performed on cladding tube material (unirradiated cold worked stress-relieved Zircaloy-4) hydrided with about 200 and 500 wppm hydrogen, and with different main hydrides orientation: either parallel or perpendicular to the circumferential tensile direction. We get the mechanical response of the material as a function of hydride orientation and hydrogen content and we investigate the deformation, damage and failure mechanisms. In both cases, digital image correlation techniques are used to estimate local and global strain distributions. Neither the tensile stress-strain response nor the global and local strain modes are significantly affected by hydrogen content or hydride orientation, but the failure modes are strongly modified. Indeed, only 200 wppm radial hydrides embrittle Zy-4: sample fail in the elastic domain at about 350 MPa before strain bands could develop; whereas in other cases sample reach at least 750 MPa before necking and final failure, in ductile or brittle mode. To model this particular heterogeneous material behavior, a non-coupled damage approach which takes into account the anisotropic distribution of the hydrides is proposed. Its parameters are identified from the macroscopic strain field measurements and a

Reactors with pressure vessel in pre-stressed concrete

International Nuclear Information System (INIS)

Devillers, Christian; Lafore, Pierre

1964-12-01

After having proposed a general description of the evolution of the general design of reactors with a vessel in pre-stressed concrete, this report outlines the interest of this technical solution of a vessel in pre-stressed concrete with integrated exchangers, which is to replace steel vessel. This solution is presented as much safer. The authors discuss the various issues related to protection: inner and outer biological protection of the vessel, material protection (against heating, steel irradiation, Wigner effect, and moderator radiolytic corrosion). They report the application of calculation methods: calculation of vessel concrete heating, study of the intermediate zone in integrated reactors, neutron spectrum and flows in the core of a graphite pile

Materials surveillance program for C-E NSSS reactor vessels

International Nuclear Information System (INIS)

Koziol, J.J.

1977-01-01

Irradiation surveillance programs for light water NSSS reactor vessels provide the means by which the utility can assess the extent of neutron-induced changes in the reactor vessel materials. These programs are conducted to verify, by direct measurement, the conservatism in the predicted radiation-induced changes and hence the operational parameters (i.e., heat-up, cooldown, and pressurization rates). In addition, such programs provide assurance that the scheduled adjustments in the operational parameters are made with ample margin for safe operation of the plant. During the past 3 years, several documents have been promulgated establishing the criteria for determining both the initial properties of the reactor vessel materials as well as measurement of changes in these initial properties as a result of irradiation. These documents, ASTM E-185-73, ''Recommended Practice for Surveillance Tests for Nuclear Reactor Vessels,'' and Appendix H to 10 CFR 50, ''Reactor Vessel Material Surveillance Program Requirements,'' are complementary to each other. They are the result of a change in the basic philosophy regarding the design and analysis of reactor vessels. In effect, the empirical ''transition temperature approach,'' which was used for design, was replaced by the ''analytical fracture mechanics approach.'' The implementation of this technique was described in Welding Research Council Bulletin 1975 and Appendix G to ASME Code Section III. Further definition of requirements appears in Appendix G to 10 CFR 50 published in July 1973. It is the intent of this paper to describe (1) a typical materials surveillance program for the reactor vessel of a Combustion Engineering NSSS, and (2) how the results of such programs, as well as experimental programs provide feed-back for improvement of materials to enhance their radiation resistance and thereby further improve the safety and reliability of future plants. (author)

Ex-vessel boiling experiments: laboratory- and reactor-scale testing of the flooded cavity concept for in-vessel core retention. Pt. II. Reactor-scale boiling experiments of the flooded cavity concept for in-vessel core retention

International Nuclear Information System (INIS)

Chu, T.Y.; Bentz, J.H.; Slezak, S.E.; Pasedag, W.F.

1997-01-01

For pt.I see ibid., p.77-88 (1997). This paper summarizes the results of a reactor-scale ex-vessel boiling experiment for assessing the flooded cavity design of the heavy water new production reactor. The simulated reactor vessel has a cylindrical diameter of 3.7 m and a torispherical bottom head. Boiling outside the reactor vessel was found to be subcooled nucleate boiling. The subcooling mainly results from the gravity head, which in turn results from flooding the side of the reactor vessel. The boiling process exhibits a cyclic pattern with four distinct phases: direct liquid-solid contact, bubble nucleation and growth, coalescence, and vapor mass dispersion. The results show that, under prototypic heat load and heat flux distributions, the flooded cavity will be effective for in-vessel core retention in the heavy water new production reactor. The results also demonstrate that the heat dissipation requirement for in-vessel core retention, for the central region of the lower head of an AP-600 advanced light water reactor, can be met with the flooded cavity design. (orig.)

An electrical pulse hydride injector (EPHI) for reactor fueling and tritium handling applications

International Nuclear Information System (INIS)

Azizov, E.A.; Kareev, Yu.A.; Savotkin, A.N.; Frunze, V.V.; Penzhorn, R.D.; Glugla, M.

1995-01-01

An electrical pulse hydride injector (EPHI) has been developed for reactor fuelling as well as for handling of hydrogen isotopes in facilities operating with tritium. Salient features of the EPHI are the accuracy with which the fuelling rate can be controlled and the avoidance of a pressurized ballast. The generator is simple and allows for safe operation with tritium. (orig.)

Pressure vessel for nuclear reactors

International Nuclear Information System (INIS)

1975-01-01

The invention applies to a pressure vessel for nuclear reactors whose shell, made of cast metal segments, has a steel liner. This liner must be constructed to withstand all operational stresses and to be easily repairable. The invention solves this problem by installing the liner at a certain distance from the inner wall of the pressure vessel shell and by filling this clearance with supporting concrete. Both the concrete and the steel liner must have a lower prestress than the pressure vessel shell. In order to avoid damage to the liner when prestressing the pressure vessel shell, special connecting elements are provided which consist of welded-on fastening elements projecting into recesses in the cast metal segments of the pressure vessel. Their design is described in detail. (TK) [de

Reactor vessel nozzle cracks: a photoelastic study

International Nuclear Information System (INIS)

Smith, C.W.

1979-01-01

A method consisting of a marriage between the ''frozen stress'' photoelastic approach and the local stress field equations of linear elastic fracture mechanics for estimating stress intensity factor distributions in three dimensional, finite cracked body problems is reviewed and extensions of the method are indicated. The method is then applied to the nuclear reactor vessel nozzle corner crack problem for both Intermediate Test Vessel and Boiling Water Reactor geometries. Results are compared with those of other investigators. 35 refs

Light Water Reactor-Pressure Vessel Surveillance project computer system

International Nuclear Information System (INIS)

Merriman, S.H.

1980-10-01

A dedicated process control computer has been implemented for regulating the metallurgical Pressure Vessel Wall Benchmark Facility (PSF) at the Oak Ridge Research Reactor. The purpose of the PSF is to provide reliable standards and methods by which to judge the radiation damage to reactor pressure vessel specimens. Benchmark data gathered from the PSF will be used to improve and standardize procedures for assessing the remaining safe operating lifetime of aging reactors. The computer system controls the pressure vessel specimen environment in the presence of gamma heating so that in-vessel conditions are simulated. Instrumented irradiation capsules, in which the specimens are housed, contain temperature sensors and electrical heaters. The computer system regulates the amount of power delivered to the electrical heaters based on the temperature distribution within the capsules. Time-temperature profiles are recorded along with reactor conditions for later correlation with specimen metallurgical changes

Power reactor pressure vessel benchmarks

International Nuclear Information System (INIS)

Rahn, F.J.

1978-01-01

A review is given of the current status of experimental and calculational benchmarks for use in understanding the radiation embrittlement effects in the pressure vessels of operating light water power reactors. The requirements of such benchmarks for application to pressure vessel dosimetry are stated. Recent developments in active and passive neutron detectors sensitive in the ranges of importance to embrittlement studies are summarized and recommendations for improvements in the benchmark are made. (author)

Analysis of aging mechanism and management for HTR-PM reactor pressure vessel

International Nuclear Information System (INIS)

Sun Yunxue; Shao Jin

2015-01-01

Reactor pressure vessel is an important part of the reactor pressure boundary, its important degree ranks high in ageing management and life assessment of nuclear power plant. Carrying out systematic aging management to ensure reactor pressure vessel keeping enough safety margins and executing design functions is one of the key factors to guarantee security and stability operation for nuclear power plant during the whole lifetime and prolong life. This paper briefly introduces the structure and aging mechanism of reactor pressure vessel in pressurized water reactor nuclear power plant, and introduces the design principle and structure characteristics of HTR-PM. At the same time, this paper carries out preliminary analysis and exploration. and discusses aging management of HTR-PM reactor pressure vessel. Finally, the advice of carring out aging management for HTR-PM reactor pressure vessel is proposed. (authors)

Neutron fluence determination for light water reactor pressure vessels

International Nuclear Information System (INIS)

Gold, R.

1994-01-01

A general description of limitations that exist in pressure vessel neutron fluence determinations for commercial light water reactors is presented. Complexity factors that arise in light water reactor pressure vessel neutron fluence calculations are identified and used to analyze calculational limitations. Two broad categories of calculational limitations are introduced, namely benchmark field limitations and deep penetration limitations. Explicit examples of limitations that can arise in each of these two broad categories are presented. These limitations are used to show that the recent draft regulatory guide for the determination of pressure vessel neutron fluence, developed by the Nuclear Regulatory Commission, is based upon procedures and assumptions that are not valid. To eliminate the complexity and limitations of calculational methods, it is recommended that the determination of light water reactor pressure vessel neutron fluence be based upon experiment. Recommendations for improved methods of pressure vessel surveillance neutron dosimetry are advanced

PWR reactor vessel in-service-inspection according to RSEM

International Nuclear Information System (INIS)

Algarotti, Marc; Dubois, Philippe; Hernandez, Luc; Landez, Jean Paul

2006-01-01

Nuclear services experience Framatome ANP (an AREVA and Siemens company) has designed and constructed 86 Pressurized Water Reactors (PWR) around the world including the three units lately commissioned at Ling Ao in the People's Republic of China and ANGRA 2 in Brazil; the company provided general and specialized outage services supporting numerous outages. Along with the American and German subsidiaries, Framatome ANP Inc. and Framatome ANP GmbH, Framatome ANP is among the world leading nuclear services providers, having experience of over 500 PWR outages on 4 continents, with current involvement in more than 50 PWR outages per year. Framatome ANP's experience in the examinations of reactor components began in the 1970's. Since then, each unit (American, French and German companies) developed automated NDT inspection systems and carried out pre-service and ISI (In-Service Inspections) using a large range of NDT techniques to comply with each utility expectations. These techniques have been validated by the utilities and the safety authorities of the countries where they were implemented. Notably Framatome ANP is fully qualified to provide full scope ISI services to satisfy ASME Section XI requirements, through automated NDE tasks including nozzle inspections, reactor vessel head inspections, steam generator inspections, pressurizer inspections and RPV (Reactor Pressure Vessel) inspections. Intercontrole (Framatome ANP subsidiary dedicated in supporting ISI) is one of the leading NDT companies in the world. Its main activity is devoted to the inspection of the reactor primary circuit in French and foreign PWR Nuclear Power Plants: the reactor vessel, the steam generators, the pressurizer, the reactor internals and reactor coolant system piping. NDT methods mastered by Intercontrole range from ultrasonic testing to eddy current and gamma ray examinations, as well as dye penetrant testing, acoustic monitoring and leak testing. To comply with the high requirements of

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

Low temperature radiation embrittlement for reactor vessel steels

International Nuclear Information System (INIS)

Ginding, I.A.; Chirkina, L.A.

1978-01-01

General conceptions of cold brittleness of bcc metals are in a review. Considered are experimental data and theoretical representations about the effect of irradiation conditions, chemical composition, phase and structural constitutions, grain size, mechanical and thermomechanical treatments on low-temperature irradiation embrittlement of reactor vessel steels. Presented are the methods for increasing radiation stability of metals (carbon and Cr-Mo steels) used in manufacturing reactor vessels

Investigation of vessel exterior air cooling for a HLMC reactor

International Nuclear Information System (INIS)

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

2000-01-01

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

Design Procedure on Stud Bolt for Reactor Vessel Assembly

International Nuclear Information System (INIS)

Kim, Jong-Wook; Lee, Gyu-Mahn; Jeoung, Kyeong-Hoon; Kim, Tae-Wan; Park, Keun-Bae; Kim, Keung-Koo

2008-10-01

The reactor pressure vessel flange is welded to the upper part of reactor pressure vessel, and there are stud holes to mount the closure head with stud bolts. The surface mating the closure head is compressed with O-ring, which acts as a sealing gasket to prevent coolant leakage. Bolted flange connections perform a very important structural role in the design of a reactor pressure vessel. Their importance stems from two important functions: (a) maintenance of the structural integrity of the connection itself, and (b) prevention of leakage through the O-ring preloaded by stud bolts. In the present study, an evaluation procedure for the design of stud bolt is developed to meet ASME code requirements. The developed design procedure could provide typical references in the development of advanced reactor design in the future

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

International Nuclear Information System (INIS)

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

2015-01-01

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

Integrity of Magnox reactor steel pressure vessels

International Nuclear Information System (INIS)

Flewitt, P.E.J.; Williams, G.H.; Wright, M.B.

1992-01-01

The background to the safety assessment of the steel reactor pressure vessels for Magnox power stations is reviewed. The evolved philosophy adopted for the 1991 safety cases prepared for the continued operation of four Magnox power stations operated by Nuclear Electric plc is described, together with different aspects of the multi-legged integrity argument. The main revisions to the materials mechanical property data are addressed together with the assessment methodology adopted and their implications for the overall integrity argument formulated for the continued safe operation of these reactor pressure vessels. (author)

A powerful methodology for reactor vessel pressurized thermal shock analysis

International Nuclear Information System (INIS)

Boucau, J.; Mager, T.

1994-01-01

The recent operating experience of the Pressurized Water Reactor (PWR) Industry has focused increasing attention on the issue of reactor vessel pressurized thermal shock (PTS). More specifically, the review of the old WWER-type of reactors (WWER 440/230) has indicated a sensitive behaviour to neutron embrittlement. This led already to some remedial actions including safety injection water preheating or vessel annealing. Such measures are usually taken based on the analysis of a selected number of conservative PTS events. Consideration of all postulated cooldown events would draw attention to the impact of operator action and control system effects on reactor vessel PTS. Westinghouse has developed a methodology which couples event sequence analysis with probabilistic fracture mechanics analyses, to identify those events that are of primary concern for reactor vessel integrity. Operating experience is utilized to aid in defining the appropriate event sequences and event frequencies of occurrence for the evaluation. Once the event sequences of concern are identified, detailed deterministic thermal-hydraulic and structural evaluations can be performed to determine the conditions required to minimize the extension of postulated flaws or enhance flaw arrest in the reactor vessel. The results of these analyses can then be used to better define further modifications in vessel and plant system design and to operating procedures. The purpose of the present paper will be to describe this methodology and to show its benefits for decision making. (author). 1 ref., 3 figs

A quantitative methodology for reactor vessel pressurized thermal shock decision making

International Nuclear Information System (INIS)

Ackerson, D.S.; Balkey, K.R.; Meyer, T.A.; Ofstun, R.P.; Rupprecht, S.D.; Sharp, D.R.

1983-01-01

The recent operating experience of the Pressurized Water Reactor (PWR) Industry has focused increasing attention on the issue of reactor vessel pressurized thermal shock (PTS). Previous reactor vessel integrity concerns have led to changes in vessel and plant system design and to operating procedures, and increased attention to the PTS issue is causing consideration of further modifications. Events such as excess feedwater, loss of normal feedwater, and steam generator tube rupture have led to significant primary system cooldowns. Each of these cooldown transients occurred concurrently with a relatively high primary system pressure. Considerations of these and other postulated cooldown events has drawn attention to the impact of operator action and control system effects on reactor vessel PTS. A methodology, which couples event sequence analysis with probabilistic fracture mechanics analyses, was developed to identify those events that are of primary concern for reactor vessel integrity. Operating experience is utilized to aid in defining the appropriate event sequences and event frequencies of occurrence for the evaluation. (orig./RW)

Stresses in transition region of VVER-1000 reactor vessels

International Nuclear Information System (INIS)

Namgung, I.; Nguye, T.L.

2014-01-01

Most of the western PWR reactor's bottom head is hemi-spherical shape, however for Russian designed VVER family of reactor it is ellipsoidal shape. The transition region from shell side to ellipsoidal head and transition top flange to cylindrical shell develop higher stress concentration than western PWR reactor vessel. This region can be modeled as conical shell with varying thickness. The theoretical derivation of stress in the thick-walled conical cylinder with varying thickness was developed and shown in detail. The results is applied to VVER-1000 reactor vessel of which shell to bottom ellipsoidal shell and shell to upper flange were investigated for stress field. The theoretical calculations were also compared with FEM solutions. An axisymmetric 3D model of VVER-1000 reactor vessel (without closure head) FEM model was created and internal hydrostatic pressure boundary condition was applied. The stress results from FEM and theoretical calculation were compared, and the discrepancies and accuracies of the theoretical results were discussed. (author)

Stresses in transition region of VVER-1000 reactor vessels

Energy Technology Data Exchange (ETDEWEB)

Namgung, I. [KEPCO International Nuclear Graduate School, Ulsan (Korea, Republic of); Nguye, T.L. [KEPCO International Nuclear Graduate School, Ulsan (Korea, Republic of); National Research Inst. of Mechanical Engineering, Hanoi City, Vietnam (China)

2014-07-01

Most of the western PWR reactor's bottom head is hemi-spherical shape, however for Russian designed VVER family of reactor it is ellipsoidal shape. The transition region from shell side to ellipsoidal head and transition top flange to cylindrical shell develop higher stress concentration than western PWR reactor vessel. This region can be modeled as conical shell with varying thickness. The theoretical derivation of stress in the thick-walled conical cylinder with varying thickness was developed and shown in detail. The results is applied to VVER-1000 reactor vessel of which shell to bottom ellipsoidal shell and shell to upper flange were investigated for stress field. The theoretical calculations were also compared with FEM solutions. An axisymmetric 3D model of VVER-1000 reactor vessel (without closure head) FEM model was created and internal hydrostatic pressure boundary condition was applied. The stress results from FEM and theoretical calculation were compared, and the discrepancies and accuracies of the theoretical results were discussed. (author)

Review of in-service thermal annealing of nuclear reactor pressure vessels

International Nuclear Information System (INIS)

Server, W.L.

1984-01-01

Radiation embrittlement of ferritic pressure vessel steels increases the ductile-brittle transition temperature and decreases the upper-shelf level of toughness as measured by Charpy impact tests. A thermal anneal cycle well above the normal operating temperature of the vessel can restore most of the original Charpy V-notch energy properties. A test reactor pressure vessel has been wet annealed at less than 343 0 C (650 0 F), and annealing of the Belgian BR-3 reactor vessel has recently taken place. An industry survey indicates that dry annealing a reactor vessel in-place is feasible, but solvable engineering problems do exist. The materials with highest radiation sensitivity in the older reactor vessels are submerged-arc weld metals with high copper and nickel concentrations. The limited Charpy V-notch and fracture toughness data available for five such welds were reviewed. The review suggested that significant recovery results from annealing at 454 0 C (850 0 F) for one week. Two of the main concerns with a localized heat treatment at 454 0 C (850 0 F) are the degree of distortion that may occur after the annealing cycle and the extent of residual stresses. A thermal and structural analysis of a reactor vessel for distortions and residual stresses found no problems with the reactor vessel itself but did indicate a rotation at the nozzle region of the vessel that would plastically deform the attached primary piping. Further analytical studies are needed. An American Society for Testing and Materials (ASTM) task group is upgrading and revising the ASTM Recommended Guide for In-Service Annealing of WaterCooled Nuclear Reactor Vessels (E 509-74) with emphasis on the materials and surveillance aspects of annealing rather than system engineering problems. System safety issues are the province of organizations other than ASTM (for example, the American Society of Mechanical Engineers Boiler and Pressure Vessel Code body)

TMI-2 reactor vessel plenum final lift

International Nuclear Information System (INIS)

Wilson, D.C.

1986-01-01

Removal of the plenum assembly from the TMI-2 reactor vessel was necessary to gain access to the core region for defueling. The plenum was lifted from the reactor vessel by the polar crane using three specially designed pendant assemblies. It was then transferred in air to the flooded deep end of the refueling canal and lowered onto a storage stand where it will remain throughout the defueling effort. The lift and transfer were successfully accomplished on May 15, 1985 in just under three hours by a lift team located in a shielded area within the reactor building. The success of the program is attributed to extensive mockup and training activities plus thorough preparations to address potential problems. 54 refs

Renovation of the sealing planes of WWER-400 reactors pressure vessel

International Nuclear Information System (INIS)

Jablonicky, P.; Pilat, P.

2007-01-01

An article describes technical solution for renovation of the sealing planes of WWER-440 reactor's pressure vessel. Four nickel sealing rings placed in four concentric grooves are providing hermetic sealing between the vessel and the lid of this type of the reactor. Impeccable seal of the reactor's pressure vessel, where the fission reaction takes place, represents a basic security factor for safe electric energy production. Principle of renovation of the reactor's pressure vessel and lid sealing planes is based on mechanical enlargement of defective grooves and following cladding of the new material by TIG welding. Final step for renovation includes machining of new grooves according to geometrical and surface quality requirements (Authors)

Design study on steam generator integration into the VVER reactor pressure vessel

International Nuclear Information System (INIS)

Hort, J.; Matal, O.

2004-01-01

The primary circuit of VVER (PWR) units is arranged into loops where the heat generated by the reactor is removed by means of main circulating pumps, loop pipelines and steam generators, all located outside the reactor pressure vessel. If the primary circuit and reactor core were integrated into one pressure vessel, as proposed, e.g., within the IRIS project (WEC), a LOCA situation would be limited by the reactor pressure vessel integrity only. The aim of this design study regarding the integration of the steam generator into the reactor pressure vessel was to identify the feasibility limits and some issues. Fuel elements and the reactor pressure vessel as used in the Temelin NPP were considered for the analysis. From among the variants analyzed, the variant with steam generators located above the core and vertically oriented circulating pumps at the RPV lower bottom seems to be very promising for future applications

In-vessel maintenance concepts for tokamak fusion reactors

International Nuclear Information System (INIS)

Kelly, V.P.; Berger, J.D.; Yount, J.A.

1983-01-01

Concepts for rail-mounted and guided in-vessel handling machines (IVM) for remote maintenance inside tokamak fusion reactors are described. The IVM designs are based on concepts for tethered remotely operated vehicles and feature the use of multiple manipulator arms for remote handling and remote-controlled TV cameras for remote viewing. The concepts include IVMs for both single or dual rail systems located in the top or bottom of the reactor vessel

Disposal of tritium-exposed metal hydrides

International Nuclear Information System (INIS)

Nobile, A.; Motyka, T.

1991-01-01

A plan has been established for disposal of tritium-exposed metal hydrides used in Savannah River Site (SRS) tritium production or Materials Test Facility (MTF) R ampersand D operations. The recommended plan assumes that the first tritium-exposed metal hydrides will be disposed of after startup of the Solid Waste Disposal Facility (SWDF) Expansion Project in 1992, and thus the plan is consistent with the new disposal requiremkents that will be in effect for the SWDF Expansion Project. Process beds containing tritium-exposed metal hydride powder will be disposed of without removal of the powder from the bed; however, disposal of tritium-exposed metal hydride powder that has been removed from its process vessel is also addressed

Shear strength of end slabs of prestressed concrete nuclear reactor vessels

International Nuclear Information System (INIS)

Reins, J.D.; Quiros, J.L. Jr.; Schnobrich, W.C.; Sozen, M.A.

1976-07-01

The report summarizes the experimental and part of the analytical work carried out in connection with an investigation of the structural strength of prestressed concrete reactor vessels. The project is part of the Prestressed Concrete Reactor Vessel Program of the Oak Ridge National Laboratory sponsored by ERDA. The objective of the current phase of the work is to develop procedures to determine the shear strength of flat end slabs of reactor vessels with penetrations

In-place thermal annealing of nuclear reactor pressure vessels

International Nuclear Information System (INIS)

Server, W.L.

1985-04-01

Radiation embrittlement of ferritic pressure vessel steels increases the ductile-brittle transition temperature and decreases the upper shelf level of toughness as measured by Charpy impact tests. A thermal anneal cycle well above the normal operating temperature of the vessel can restore most of the original Charpy V-notch energy properties. The Amry SM-1A test reactor vessel was wet annealed in 1967 at less than 343 0 C (650 0 F), and wet annealing of the Belgian BR-3 reactor vessel at 343 0 C (650 0 F) has recently taken place. An industry survey indicates that dry annealing a reactor vessel in-place at temperatures as high as 454 0 C (850 0 F) is feasible, but solvable engineering problems do exist. Economic considerations have not been totally evaluated in assessing the cost-effectiveness of in-place annealing of commercial nuclear vessels. An American Society for Testing and Materials (ASTM) task group is upgrading and revising guide ASTM E 509-74 with emphasis on the materials and surveillance aspects of annealing rather than system engineering problems. System safety issues are the province of organizations other than ASTM (e.g., the American Society of Mechanical Engineers Boiler and Pressure Vessel Code body)

Neutron Fluence And DPA Rate Analysis In Pebble-Bed HTR Reactor Vessel Using MCNP

Science.gov (United States)

Hamzah, Amir; Suwoto; Rohanda, Anis; Adrial, Hery; Bakhri, Syaiful; Sunaryo, Geni Rina

2018-02-01

In the Pebble-bed HTR reactor, the distance between the core and the reactor vessel is very close and the media inside are carbon and He gas. Neutron moderation capability of graphite material is theoretically lower than that of water-moderated reactors. Thus, it is estimated much more the fast neutrons will reach the reactor vessel. The fast neutron collisions with the atoms in the reactor vessel will result in radiation damage and could be reducing the vessel life. The purpose of this study was to obtain the magnitude of neutron fluence in the Pebble-bed HTR reactor vessel. Neutron fluence calculations in the pebble-bed HTR reactor vessel were performed using the MCNP computer program. By determining the tally position, it can be calculated flux, spectrum and neutron fluence in the position of Pebble-bed HTR reactor vessel. The calculations results of total neutron flux and fast neutron flux in the reactor vessel of 1.82x108 n/cm2/s and 1.79x108 n/cm2/s respectively. The fast neutron fluence in the reactor vessel is 3.4x1017 n/cm2 for 60 years reactor operation. Radiation damage in stainless steel material caused by high-energy neutrons (> 1.0 MeV) will occur when it has reached the neutron flux level of 1.0x1024 n/cm2. The neutron fluence results show that there is no radiation damage in the Pebble-bed HTR reactor vessel, so it is predicted that it will be safe to operate at least for 60 years.

The analysis of reactor vessel surveillance program data

International Nuclear Information System (INIS)

Norris, E.B.

1979-01-01

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

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

International Nuclear Information System (INIS)

1997-08-01

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

In-Vessel Retention via External Reactor Cooling

Energy Technology Data Exchange (ETDEWEB)

Bachrata, Andrea [CTU in Prague, Faculty of nuclear sciences and physical engineering, V Holesovickach 2 180 00, Prague 8 (Czech republic)

2008-07-01

In-vessel (corium) retention (IVR) via external reactor pressure vessel (RPV) cooling is considered to be an effective severe accident management strategy for corium localisation and stabilisation. The main idea of IVR strategy consists in flooding the reactor cavity and transferring the decay heat through the wall of RPV to the recirculating water and than to the atmosphere of the containment of nuclear power plant. The aim of this strategy is to localise and to stabilise the corium inside the RPV. Not using this procedure could destroy the integrity of RPV and might cause the interaction of the corium with the concrete at the bed of the reactor cavity. Several experimental facilities and computer codes (MVITA, ASTEC module DIVA and CFD codes) were applied to simulate the IVR strategy for concrete reactor designs. The necessary technical modifications concerning the implementation of IVR concept were applied at the Loviisa NPP (VVER-440/V213). This strategy is also an important part of the advanced reactor designs AP600 and AP1000. (authors)

Burnup influence on the VVER-1000 reactor vessel neutron fluence evaluation

International Nuclear Information System (INIS)

Panayotov, I.; Mihaylov, N.; Ilieva, K.; Kirilova, D.; Manolova, M.

2009-01-01

The neutron fluence of the vessels of the reactors is determined regularly accordingly the RPV Surveillance Program of the Kozloduy NPP Unit 5 and 6 in order to assess the state of the metal vessel and their radiation damaging. The calculations are carried out by the method of discrete ordinates used in the TORT program for operated reactor cycles. An average reactor spectrum corresponding to fresh U-235 fuel is used as an input neutron source. The impact of the burn up of the fuel on the neutron fluence of VVER-1000 reactor vessel is evaluated. The calculations of isotopic concentrations of U-235 and Pu-239 corresponding to 4 years burn up were performed by the module SAS2H of the code system SCALE 4.4. Since fresh fuel or 4 years burn up fuel assembly are placed in periphery of reactor core the contribution of Pu-239 of first year burn up and of 4 years burn up is taken in consideration. Calculations of neutron fluence were performed with neutron spectrum for fresh fuel, for 1 year and for 4 years burn up fuel. Correction factors for neutron fluence at the inner surface of the reactor vessel, in 1/4 depth of the vessel and in the air behind the vessel were obtained. The correction coefficient could be used when the neutron fluence is assessed so in verification when the measured activity of ex-vessel detectors is compared with calculated ones. (authors)

Burnup influence on the WWER-1000 reactor vessel neutron fluence evaluation

International Nuclear Information System (INIS)

Panayotov, I.; Mihaylov, N.; Ilieva, K.; Kirilova, D.; Manolova, M.

2009-01-01

The neutron fluence of the vessels of the reactors is determined regularly accordingly the RPV Surveillance Program of Kozloduy NPP Unit 5 and 6 in order to assess the state of the metal vessel and their radiation damaging. The calculations are carried out by the method of discrete ordinates used in the TORT program for operated reactor cycles. An average reactor spectrum corresponding to fresh U-235 fuel is used as an input neutron source. The impact of the burn up of the fuel on the neutron fluence of WWER-1000 reactor vessel is evaluated. The calculations of isotopic concentrations of U-235 and Pu-239 corresponding to 4 years burn up were performed by the module SAS2H of the code system SCALE 4.4. Since fresh fuel or 4 years burn up fuel assembly are placed in periphery of reactor core the contribution of Pu-239 of first year burn up and of 4 years burn up is taken in consideration. Calculations of neutron fluence were performed with neutron spectrum for fresh fuel, for 1 year and for 4 years burn up fuel. Correction factors for neutron fluence at the inner surface of the reactor vessel, in ? depth of the vessel and in the air behind the vessel were obtained. The correction coefficient could be used when the neutron fluence is assessed so in verification when the measured activity of ex-vessel detectors is compared with calculated ones. (Authors)

Surveillance specimen programmes for WWER reactor vessels in the Czech Republic

International Nuclear Information System (INIS)

Brynda, J.; Hogel, J.; Brumovsky, M.

2003-01-01

The present state of materials degradation in WWER reactor pressure vessels manufactured in the Czech Republic is highlighted. The standard surveillance program for WWER-440/V-213 type reactors is described and its deficiencies together with the main results obtained are discussed. A new supplementary surveillance program meeting all requirements for PWR type reactors has been developed and launched. An entirely new design was chosen for the surveillance programme for WWER-1000/V-320 type reactor pressure vessels. Materials selection, container design and location as well as the withdrawal plan connected with ex-vessel fluence monitoring are described

Reactor vessel stud tensioner

International Nuclear Information System (INIS)

Malandra, L.J.; Beer, R.W.; Salton, R.B.; Spiegelman, S.R.; Cognevich, M.L.

1982-01-01

A quick-acting stud tensioner, for facilitating the loosening or tightening of a stud nut on a reactor vessel stud, has gripper jaws which when the tensioner is lowered into engagement with the upper end of the stud are moved inwards to grip the upper end and which when the tensioner is lifted move outward to release the upper end. (author)

Elements of thought on corium containment strategy in reactor vessel

International Nuclear Information System (INIS)

2015-01-01

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

Head spray nozzle in reactor pressure vessel

International Nuclear Information System (INIS)

Hatano, Shun-ichi.

1990-01-01

In a reactor pressure vessel of a BWR type reactor, a head spray nozzle is used for cooling the head of the pressure vessel and, in view of the thermal stresses, it is desirable that cooling is applied as uniformly as possible. A conventional head spray is constituted by combining full cone type nozzles. Since the sprayed water is flown down upon water spraying and the sprayed water in the vertical direction is overlapped, the flow rate distribution has a high sharpness to form a shape as having a maximum value near the center and it is difficult to obtain a uniform flow rate distribution in the circumferential direction. Then, in the present invention, flat nozzles each having a spray water cross section of laterally long shape, having less sharpness in the circumferential distribution upon spraying water to the inner wall of the pressure vessel and having a wide angle of water spray are combined, to make the flow rate distribution of spray water uniform in the inner wall of the pressure vessel. Accordingly, the pressure vessel can be cooled uniformly and thermal stresses upon cooling can be decreased. (N.H.)

Thermal embrittlement of reactor vessel steels

International Nuclear Information System (INIS)

Corwin, W.R.; Nanstad, R.K.; Alexander, D.J.; Stoller, R.E.; Wang, J.A.; Odette, G.R.

1995-01-01

As a result of observations of possible thermal embrittlement from recent studies with welds removed from retired steam generators of the Palisades Nuclear Plant (PNP), an assessment was made of thermal aging of reactor pressure vessel (RPV) steels under nominal reactor operating conditions. Discussions are presented on (1) data from the literature regarding relatively low-temperature thermal embrittlement of RPV steels; (2)relevant data from the US power reactor-embrittlement data base (PR-EDB); and (3)potential mechanisms of thermal embrittlement in low-alloy steels

Provision of reliable core cooling in vessel-type boiling reactors

International Nuclear Information System (INIS)

Alferov, N.S.; Balunov, B.F.; Davydov, S.A.

1987-01-01

Methods for providing reliable core cooling in vessel-type boiling reactors with natural circulation for heat supply are analysed. The solution of this problem is reduced to satisfaction of two conditions such as: water confinement over the reactor core necessary in case of an accident and confinement of sufficient coolant flow rate through the bottom cross section of fuel assemblies for some time. The reliable fuel element cooling under conditions of a maximum credible accident (brittle failure of a reactor vessel) is shown to be provided practically in any accident, using the safety vessel in combination with the application of means of standard operation and minimal composition and capacity of ECCS

Design of the prestressed concrete reactor vessel for gas-cooled heating reactors

International Nuclear Information System (INIS)

Becker, G.; Notheisen, C.; Steffen, G.

1987-01-01

The GHR pebble bed reactor offers a simple, safe and economic possibility of heat generation. An essential component of this concept is the prestressed concrete reactor vessel. A system of cooling pipes welded to the outer surface of the liner is used to transfer the heat from the reactor to the intermediate circuit. The high safety of this vessel concept results from the clear separation of the functions of the individual components and from the design principle of the prestressed conncrete. The prestressed concrete structure is so designed that failure can be reliably ruled out under all operating and accident conditions. Even in the extremely improbable event of failure of all decay heat removal systems when decay heat and accumulated heat are transferred passively by natural convection only, the integrity of the vessel remains intact. For reasons of plant availability the liner and the liner cooling system shall be designed so as to ensure safe elimination of failure over the total operating life. The calculations which were peformed partly on the basis of extremely adverse assumption, also resulted in very low loads. The prestressed concrete vessel is prefabricated to the greatest possible extent. Thus a high quality and optimized fabrication technology can be achieved especially for the liner and the liner cooling system. (orig./HP)

Embrittlement of the nuclear icebreaker Lenin reactor pressure vessel materials reconstruction

International Nuclear Information System (INIS)

Krasikov, E.A.; Nikolaenko, V.A.

2008-01-01

Paper deals with the results of the efforts to examine the radiation damage of the Lenin nuclear-powered ice-breaker decommissioned reactor pressure vessel on the basis of which one has determined the peculiar features of the metal radiation embrittlement. Under 10 10 -10 11 s -1 cm -2 low density neutron flux irradiation one notes the most intensive embrittlement of the metal. Then, as the noxious element content in the metal matrix grows smaller the embrittlement reduces up to the change of sign as to the normal curve plotted at the neutron flux density exceeding 10 13 s -1 cm -2 . One assumes that as a result of the low density neutron flux irradiation the reactor pressure vessel edge spaces at some operation stages may be damaged more severely in contrast to these near the reactor core. The neutron irradiation density is the factor affecting the reactor vessel material embrittlement, that is why, it is important to study the damage mechanism of the materials of the power reactor vessels under design characterized by the low radiation load. The mentioned is important, as well, to evaluate the efficiency of the efforts undertaken to mitigate the effect of the neutron radiation on the reactor vessel [ru

TMI-2 reactor vessel head removal

International Nuclear Information System (INIS)

Bengel, P.R.; Smith, M.D.; Estabrook, G.A.

1984-12-01

This report describes the safe removal and storage of the Three Mile Island Unit 2 reactor vessel head. The head was removed in July 1984 to permit the removal of the plenum and the reactor core, which were damaged during the 1979 accident. From July 1982, plans and preparations were made using a standard head removal procedure modified by the necessary precautions and changes to account for conditions caused by the accident. After data acquisition, equipment and structure modifications, and training the head was safely removed and stored and the internals indexing fixture and a work platform were installed on top of the vessel. Dose rates during and after the operation were lower than expected; lessons were learned from the operation which will be applied to the continuing fuel removal operations activities

TMI-2 reactor vessel head removal

International Nuclear Information System (INIS)

Bengel, P.R.; Smith, M.D.; Estabrook, G.A.

1985-09-01

This report describes the safe removal and storage of the Three Mile Island Unit 2 (TMI-2) reactor vessel head. The head was removed in July 1984 to permit the removal of the plenum and the reactor core, which were damaged during the 1979 accident. From July 1982, plans and preparations were made using a standard head removal procedure modified by the necessary precautions and changes to account for conditions caused by the accident. After data acquisition, equipment and structure modifications, and training, the head was safely removed and stored; and the internals indexing fixture and a work platform were installed on top of the vessel. Dose rates during and after the operation were lower than expected; lessons were learned from the operation which will be applied to the continuing fuel removal operations activities

Leak detector for reactor pressure vessel

International Nuclear Information System (INIS)

Morimoto, Mikio.

1991-01-01

A branched pipe is disposed to a leak off pipeline led from a flange surface which connects the main body and the upper lid of a reactor pressure vessel. An exhaust pump is disposed to the branched pipe and a moisture gage is disposed on the side of the exhaustion and a dry air supplier is connected to the branched pipe. Upon conducting a pressure-proof leak test for the reactor pressure vessel, the exhaust pump is operated and an electromagnet valve disposed at the upstream of the dry air supplier is opened and closed repeatedly. The humidity of air sucked by the exhaust pump is detected by the moisture gage. If leaks should be caused in the joining surface of the flange, leaked water is diffused as steams. Accordingly, occurrence of leak can be detected instantly based on the comparison with the moisture level of the dry air as a standard. In this way, a leak test can be conducted reliably in a short period of time with no change of for the reactor pressure container itself. (I.N.)

Damage by radiation in structural materials of BWR reactor vessels

International Nuclear Information System (INIS)

Robles, E.; Balcazar, M.; Alpizar, A.M.; Calderon, B.E.

2002-01-01

The structural materials which are manufactured the pressure vessels of the BWR reactors undergo degradation in their mechanical properties mainly due to the damage produced by the fast neutrons (E> 1 MeV) coming from the reactor core. The mechanisms of neutron damage in this type of materials are experimentally studied, through the irradiation of vessel steel in experimental reactors for a quickly ageing. Alternately the neutron damage through steel irradiation with heavy ions is simulated. In this work the first results of the damage induced by irradiation of a similar steel to the vessel of a BWR reactor are shown. The irradiation was performed with fast neutrons (E> 1 MeV, fluence of 1.45 x 10 18 n/cm 2 ) in the TRIGA Mark III Salazar reactor and separately with Ni +3 ions in a Tandetrom accelerator (E= 4.8 MeV and an ion flux rank of 0.1 to 53 ions/A 2 ). (Author)

Studies of fragileness in steels of vessels of BWR reactors

International Nuclear Information System (INIS)

Robles, E.F.; Balcazar, M.; Alpizar, A.M.; Calderon, B.E.

2003-01-01

The structural materials with those that are manufactured the pressure vessels of the BWR reactors, suffer degradation in its mechanical properties mainly to the damage taken place by the fast neutrons (E > 1 MeV) coming from the reactor core. Its are experimentally studied those mechanisms of neutron damage in this material type, by means of the irradiation of steel vessel in experimental reactors to age them quickly. Alternatively it is simulated the neutron damage by means of irradiation of steel with heavy ions. In this work those are shown first results of the damage induced by irradiation from a similar steel to the vessel of a BWR reactor. The irradiation was carried out with fast neutrons (E > 1 MeV, fluence of 1.45 x 10 18 n/cm 2 ) in the TRIGA MARK lll reactor and separately with Ni +3 ions in a Tandetrom accelerator, E = 4.8 MeV and range of the ionic flow of 0.1 to 53 iones/A 2 . (Author)

A wall-crawling robot for reactor vessel inspection in advanced reactors

International Nuclear Information System (INIS)

Spelt, P.F.; Crane, C.; Feng, L.; Abidi, M.; Tosunoglu, S.

1994-01-01

A consortium of four universities and the Center for Engineering Systems Advanced Research of the Oak Ridge National Laboratory has designed a prototype wall-crawling robot to perform weld inspection in advanced nuclear reactors. Design efforts for the reactor vessel inspection robot (RVIR) concentrated on the Advanced Liquid Metal Reactor because it presents the most demanding environment in which such a robot must operate. The RVIR consists of a chassis containing two sets of suction cups that can alternately grasp the side of the vessel being inspected, providing both locomotion and steering functions. Sensors include three CCD cameras and a weld inspection device based on new shear-wave technology. The restrictions of the inspection environment presented major challenges to the team. These challenges were met in the prototype, which has been tested in a non-radiation, room-temperature mockup of the robot work environment and shown to perform as expected. (author)

A wall-crawling robot for reactor vessel inspection in advanced reactors

International Nuclear Information System (INIS)

Spelt, P.F.; Crane, C.; Feng, L.; Abidi, M.; Tosunoglu, S.

1994-01-01

A consortium of four universities and the Center for Engineering Systems Advanced Research of the Oak Ridge National Laboratory has designed a prototype wall-crawling robot to perform weld inspection in advanced nuclear reactors. Design efforts for the reactor vessel inspection robot (RVIR) concentrated on the Advanced Liquid Metal Reactor because it presents the most demanding environment in which such a robot must operate. The RVIR consists of a chassis containing two sets of suction cups that can alternately grasp the side of the vessel being inspected, providing both locomotion and steering functions. Sensors include three CCD cameras and a weld inspection device based on new shear-wave technology. The restrictions of the inspection environment presented major challenges to the team. These challenges were met in the prototype, which has been tested in a non-radiation, room-temperature mockup of the robot work environment and shown to perform as expected

Modeling irradiation embrittlement in reactor pressure vessel steels

International Nuclear Information System (INIS)

Odette, G.R.

1998-01-01

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

Research and development of the prestressed concrete reactor vessel

International Nuclear Information System (INIS)

Shiozawa, Shoji; Omata, Ippei; Nakamura, Norio

1975-01-01

Compared with the steel reactor vessel, the prestressed concrete reactor vessel (PCRV) is said to be superior in safety and economy. One of the characteristics of the high temperature gas cooled reactor (HTGR) is the adoption of the PCRV instead of the steel reactor vessel to ensure safety. In order to improve safety characteristics, it is necessary for the PCRV to be provided with more reliable functions. When the multi-purpose HTGR or the gas cooled fast breeder reactor (GCFR) are realized in future, more severe conditions of technology will be imposed on the PCRV, and accordingly, technical developments are now increasingly required. IHI is now proceeding with the technical research and development on the PCRV, in which a basic study of its liner cooling system has already been completed. In this study applying a large cylindrical PCRV model, comparison was made between experimental data and analyses concerning the liner cooling system, and the results of analytical technique have been evaluated. The analytical technique established this time is applicable to the estimation of temperature distribution in the concrete of a large PCRV and also to the evaluation of the liner cooling system. (auth.)

Reactor pressure vessel steels

International Nuclear Information System (INIS)

Van De Velde, J.; Fabry, A.; Van Walle, E.; Chaouuadi, R.

1998-01-01

Research and development activities related to reactor pressure vessel steels during 1997 are reported. The objectives of activities of the Belgian Nuclear Research Centre SCK/CEN in this domain are: (1) to develop enhanced surveillance concepts by applying micromechanics and fracture-toughness tests to small specimens, and by performing damage modelling and microstructure characterization; (2) to demonstrate a methodology on a broad database; (3) to achieve regulatory acceptance and industrial use

Control Rod Drive Mechanism Installed in the Internal of Reactor Pressure Vessel

Energy Technology Data Exchange (ETDEWEB)

Choi, M. H.; Choi, S.; Park, J. S.; Lee, J. S.; Kim, D. O.; Hur, N. S.; Hur, H.; Yu, J. Y

2008-09-15

This report describes the review results and important technologies related to the in-vessel type control rod drive mechanism. Generally, most of the CRDMs used in the PWR are attached outside of the reactor pressure vessel, and the pernetration of the vessel head can not avoid. However, in-vessel type CRDMs, which are installed inside the reactor vessel, can eliminate the possibility of rod ejection accidents and the penetration of the vessel head, and provide a compact design of the reactor vessel and containment. There are two kinds of in-vessel type CRDM concerning the driving force-driven by a driving motor and by a hydraulic force. Motor driven CRDMs have been mainly investigated in Japan(MRX, IMR, DRX, next generation BWR etc.), and developed the key components such as a canned motor, an integrated rod position indicator, a separating ball-nut and a ball bearing that can operate under the water conditions of a high temperature and pressure. The concept of hydraulically driven CRDMs have been first reported by KWU and Siemens for KWU 200 reactor, and Argentina(CAREM) and China(NHR-5, NHR-200) have been developed the internal CRDM with the piston and cylinder of slightly different geometries. These systems are driven by the hydraulic force which is produced by pumps outside of the reactor vessel and transmitted through a pipe penetrating the reactor vessel, and needs complicated control and piping systems including pumps, valves and pipes etc.. IRIS has been recently decided the internal CRDMs as the reference design, and an analytical and experimental investigations of the hydraulic drive concept are performed by POLIMI in Italy. Also, a small French company, MP98 has been developed a new type of control rods, called 'liquid control rods', where reactivity is controlled by the movement of a liquid absorber in a manometer type device.

Shock loading of reactor vessel following hypothetical core disruptive accident

International Nuclear Information System (INIS)

Srinivas, G.; Doshi, J.B.

1990-01-01

Hypothetical Core Disruptive Accident (HCDA) has been historically considered as the maximum credible accident in Fast Breeder Reactor systems. Environmental consequences of such an accident depends to a great extent on the ability of the reactor vessel to maintain integrity during the shock loading following an HCDA. In the present paper, a computational model of the reactor core and the surrounding coolant with a free surface is numerical technique. The equations for conservation of mass, momentum and energy along with an equation of state are considered in two dimensional cylindrical geometry. The reactor core at the end of HCDA is taken as a bubble of hot, vaporized fuel at high temperature and pressure, formed at the center of the reactor vessel and expanding against the surrounding liquid sodium coolant. The free surface of sodium at the top of the vessel and the movement of the core bubble-liquid coolant interface are tracked by Marker and Cell (MAC) procedure. The results are obtained for the transient pressure at the vessel wall and also for the loading on the roof plug by the impact of the slug of liquid sodium. The computer code developed is validated against a benchmark experiment chosen to be ISPRA experiment reported in literature. The computer code is next applied to predict the loading on the Indian Prototype Fast Breeder Reactor (PFBR) being developed at Kalpakkam

Construction of reactor vessel bottom of prestressed reinforced concrete

International Nuclear Information System (INIS)

Sitnikov, M.I.; Metel'skij, V.P.

1980-01-01

Methods are described for building reactor vessel bottoms of prestressed reinforced concrete during NPPs construction in Great Britain, France, Germany (F.R.) and the USA. Schematic of operations performed in succession is presented. Considered are different versions of one of the methods for concreting a space under a facing by forcing concrete through a hole in the facing. The method provides tight sticking of the facing to the reactor vessel bottom concrete

After-operating properties of nuclear reactor vessel materials of Lenin atomic ice breaker and prospective of reactor vessels radiation life prolongation

International Nuclear Information System (INIS)

Platonov, P.A.; Shtrombakh, Ya.I.; Amaev, A.D.; Krasikov, E.A.; Korolev, Yu.N.; Zabusov, O.O.; Glushakov, G.M.

2001-01-01

A post-operational state of the icebreaker Lenin reactor vessel metal is investigated. It is shown that a base metal of the icebreaker Lenin reactor vessel is of high quality as by an initial value of critical temperature of embrittlement, so by its radiation resistance. The weld metal possesses a sufficient radiation resistance but has an insufficient initial ductile-brittle transition temperature (approximately 63 Deg C). It is necessary to note that the final stage of operation for nuclear steam-generating plant should be carried out at the coolant temperature as high as possible [ru

External Reactor Vessel Cooling Evaluation for Severe Accident Mitigation in NPP Krsko

International Nuclear Information System (INIS)

Mihalina, M.; Spalj, S.; Glaser, B.

2016-01-01

The In-Vessel corium Retention (IVR) through the External Reactor Vessel Cooling (ERVC) is mean for maintaining the reactor vessel integrity during a severe accident, by cooling and retaining the molten material inside the reactor vessel. By doing this, significant portion of severe accident negative phenomena connected with reactor vessel failure could be avoided. In this paper, analysis of NPP Krsko applicability for IVR strategy was performed. It includes overview of performed plant related analysis with emphasis on wet cavity modification, plant's site specific walk downs, new applicable probabilistic and deterministic analysis, evaluation of new possibilities for ERVC strategy implementation regarding plant's post-Fukushima improvements and adequacy with plant's procedures for severe accident mitigation. Conclusion is that NPP Krsko could perform in-vessel core retention by applying external reactor vessel cooling strategy with reasonable confidence in success. Per probabilistic and deterministic analysis, time window for successful ERVC strategy performance for most dominating plant damage state scenarios is 2.5 hours, when onset of core damage is observed. This action should be performed early after transition to Severe Accident Management Guidance's (SAMG). For loss of all AC power scenario, containment flooding could be initiated before onset of core damage within related emergency procedure. To perform external reactor vessel cooling, reactor water storage tank gravity drain with addition of alternate water is needed to be injected into the containment. ERVC strategy will positively interfere with other severe accident strategies. There are no negative effects due to ERVC performance. New flooding level will not threaten equipment and instrumentation needed for long term SAMGs performance and eventually diluted containment sump borated water inventory will not cause return to criticality during eventual recirculation phase due to the

System for cooling the containment vessel of a nuclear reactor

International Nuclear Information System (INIS)

Costes, Didier.

1982-01-01

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

Structural integrity evaluation of PWR nuclear reactor pressure vessels

International Nuclear Information System (INIS)

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

1999-01-01

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

Segmentation and packaging reactor vessels internals

International Nuclear Information System (INIS)

Boucau, Joseph

2014-01-01

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

Limiting Factors for External Reactor Vessel Cooling

International Nuclear Information System (INIS)

Cheung, F.B.

2005-01-01

The method of external reactor vessel cooling (ERVC) that involves flooding of the reactor cavity during a severe accident has been considered a viable means for in-vessel retention (IVR). For high-power reactors, however, there are some limiting factors that might adversely affect the feasibility of using ERVC as a means for IVR. In this paper, the key limiting factors for ERVC have been identified and critically discussed. These factors include the choking limit for steam venting (CLSV) through the bottleneck of the vessel/insulation structure, the critical heat flux (CHF) for downward-facing boiling on the vessel outer surface, and the two-phase flow instabilities in the natural circulation loop within the flooded cavity. To enhance ERVC, it is necessary to eliminate or relax these limiting factors. Accordingly, methods to enhance ERVC and thus improve margins for IVR have been proposed and demonstrated, using the APR1400 as an example. The strategy is based on using two distinctly different methods to enhance ERVC. One involves the use of an enhanced vessel/insulation design to facilitate steam venting through the bottleneck of the annular channel. The other involves the use of an appropriate vessel coating to promote downward-facing boiling. It is found that the use of an enhanced vessel/insulation design with bottleneck enlargement could greatly facilitate the process of steam venting through the bottleneck region as well as streamline the resulting two-phase motions in the annular channel. By selecting a suitable enhanced vessel/insulation design, not only the CLSV but also the CHF limits could be significantly increased. In addition, the problem associated with two-phase flow instabilities and flow-induced mechanical vibration could be minimized. It is also found that the use of vessel coatings made of microporous metallic layers could greatly facilitate downward-facing boiling on the vessel outer surface. With vessel coatings, the local CHF limits at

Verification of codes used for the nuclear safety assessment of the small space heterogeneous reactors with zirconium hydride moderator

International Nuclear Information System (INIS)

Glushkov, E.S.; Gomin, E.A.; Kompaniets, G.V.

1994-01-01

Computer codes used for assessment of nuclear safety for space NPP are compared taking as an example small-sized heterogeneous reactor with zirconium hydride moderator of the Topaz-2 facility. The code verifications are made for five different variants

Safety vessels for explosive fusion reactor

International Nuclear Information System (INIS)

Mineev, V.

1994-01-01

The failure of several types of geometrically similar cylindrical and spherical steel and glass fibers vessels filled with water or air was investigated when an explosive charge of TNT was detonated in the center. Vessels had radius 50-1000 mm, thickness of walls 2-20%. The detonation on TNT imitated energy release. The parameter: K = M/mf is a measure of the strength of the vessel where M is the mass of the vessel, and mf is the mass of TNT for which the vessel fails. This demanded 2-4 destroyed and nondestroyed shots. It may be showed that: K=A/σ f where σ f is the fracture stress of the material vessel, and A = const = F(energy TNT, characteristic of elasticity of vessel material). The chief results are the following: (1) A similar increase in the geometrical dimensions of steel vessels by a factor of 10 leads to the increase of parameter K in about 5 times and to decrease of failure deformation in 7 times (scale effect). (2) For glass fibers, scale effect is absent. (3) This problem is solved in terms of theory energetic scale effect. (4) The concept of TNT equivalent explosive makes it possible to use these investigations to evaluate the response of safety vessels for explosive fusion reactor

Hydride formation on deformation twin in zirconium alloy

Energy Technology Data Exchange (ETDEWEB)

Kim, Ju-Seong [Korea Atomic Energy Research Institute, 989-111 Daedeokdaero, Yuseong-gu, Daejeon, 305-353 (Korea, Republic of); Kim, Sung-Dae [Korea Institute of Material Science (KIMS), 797 Changwondaero, Changwon, Gyeongnam, 642-831 (Korea, Republic of); Yoon, Jonghun, E-mail: yooncsmd@gmail.com [Department of Mechanical Engineering, Hanyang University, 1271 Sa3-dong, Sangrok-gu, Ansan-si, Gyeonggi-do, 426-791 (Korea, Republic of)

2016-12-15

Hydrides deteriorate the mechanical properties of zirconium (Zr) alloys used in nuclear reactors. Intergranular hydrides that form along grain boundaries have been extensively studied due to their detrimental effects on cracking. However, it has been little concerns on formation of Zr hydrides correlated with deformation twins which is distinctive heterogeneous nucleation site in hexagonal close-packed metals. In this paper, the heterogeneous precipitation of Zr hydrides at the twin boundaries was visualized using transmission electron microscopy. It demonstrates that intragranular hydrides in the twinned region precipitates on the rotated habit plane by the twinning and intergranular hydrides precipitate along the coherent low energy twin boundaries independent of the conventional habit planes. Interestingly, dislocations around the twin boundaries play a substantial role in the nucleation of Zr hydrides by reducing the misfit strain energy.

Analysis of nuclear reactor pressure vessel flanges

International Nuclear Information System (INIS)

Oliveira, C.A.N. de; Augusto, O.B.

1985-01-01

This work proposes a methodology for the structural analysis of high diameter nuclear reactor pressure vessel flanges. In the analysis the vessel is divided into shell-of-revolution elements, the flanges are represented by rigid rings, and the bolts are treated as beams. The flexibility method is used for solving the problem, and the results are compared with results obtained by the finite element method. (Author) [pt

Advanced Approach of Reactor Pressure Vessel In-service Inspection

International Nuclear Information System (INIS)

Matokovic, A.; Picek, E.; Pajnic, M.

2006-01-01

The most important task of every utility operating a nuclear power plant is the continuously keeping of the desired safety and reliability level. This is achieved by the performance of numerous inspections of the components, equipment and system of the nuclear power plant in operation and in particular during the scheduled maintenance periods at re-fueling time. Periodic non-destructive in-service inspections provide most relevant criteria of the integrity of primary circuit pressure components. The task is to reliably detect defects and realistically size and characterize them. One of most important and the most extensive examination is a reactor pressure vessel in-service inspection. That inspection demand high standards of technology and quality and continual innovation in the field of non-destructive testing (NDT) advanced technology as well as regarding reactor pressure vessel tool and control systems. A remote underwater contact ultrasonic technique is employed for the examination of the defined sections (reactor welds), whence eddy current method is applied for clad surface examinations. Visual inspection is used for examination of the vessel inner surface. The movement of probes and data positioning are assured by using new reactor pressure vessel tool concept that is fully integrated with NDT systems. The successful performance is attributed thorough pre-outage planning, training and successful performance demonstration qualification of chosen NDT techniques on the specimens with artificial and/or real defects. Furthermore, use of advanced approach of inspection through implementation the state of the art examination equipment significantly reduced the inspection time, radiation exposure to examination personnel, shortening nuclear power plant outage and cutting the total inspection costs. The advanced approach as presented in this paper offer more flexibility of application (non-destructive tests, local grinding action as well as taking of boat samples

In-service ultrasonic inspection of nuclear reactor pressure vessels

International Nuclear Information System (INIS)

Prepechal, J.; Sulc, J.

1982-01-01

Ultrasonic tests of pressure vessels for WWER 440 reactors, type 213 V, are carried out partly manually and partly by test equipment. The inner surface of the pressure vessel is tested using device REACTORTEST TRC which is fully mobile. The outer surface of the cylindrical parts and bottoms of the body is tested using handling equipment permanently in-built under the pressure vessel and dismountable testing heads. A set of these heads may be used for two reactor units. The testing equipment REACTORTEST TRC is equipped with a TRC 800 ultrasound device. The equipment for testing the outer surface of the vessel operates with the UDAR 16 ultrasound apparatus to which may be simultaneously connected 10 ultrasound probes and six probes for acoustic feedback. The whole system of ultrasonic tests makes possible a first-rate and reliable volume control of the whole pressure vessel and all points where cracks may originate and grow. (Z.M.)

Design and analysis of multicavity prestressed concrete reactor vessels

International Nuclear Information System (INIS)

Goodpasture, D.W.; Burdette, E.G.; Callahan, J.P.

1977-01-01

During the past 25 years, a rather rapid evolution has taken place in the design and use of prestressed concrete reactor vessels (PCRVs). Initially the concrete vessel served as a one-to-one replacement for its steel counterpart. This was followed by the development of the integral design which led eventually to the more recent multicavity vessel concept. Although this evolution has seen problems in construction and operation, a state-of-the-art review which was recently conducted by the Oak Ridge National Laboratory indicated that the PCRV has proven to be a satisfactory and inherently safe type of vessel for containment of gas-cooled reactors from a purely functional standpoint. However, functionalism is not the only consideration in a demanding and highly competitive industry. A summary is presented of the important considerations in the design and analysis of multicavity PCRVs together with overall conclusions concerning the state of the art of these vessels

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

International Nuclear Information System (INIS)

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

2007-01-01

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

Liquid metal systems development: reactor vessel support structure evaluation

International Nuclear Information System (INIS)

McEdwards, J.A.

1981-01-01

Results of an evaluation of support structures for the reactor vessel are reported. The U ring, box ring, integral ring, tee ring and tangential beam supports were investigated. The U ring is the recommended vessel support structure configuration

The Effect of Hydrogen and Hydrides on the Integrity of Zirconium Alloy Components Delayed Hydride Cracking

CERN Document Server

Puls, Manfred P

2012-01-01

By drawing together the current theoretical and experimental understanding of the phenomena of delayed hydride cracking (DHC) in zirconium alloys, The Effect of Hydrogen and Hydrides on the Integrity of Zirconium Alloy Components: Delayed Hydride Cracking provides a detailed explanation focusing on the properties of hydrogen and hydrides in these alloys. Whilst the focus lies on zirconium alloys, the combination of both the empirical and mechanistic approaches creates a solid understanding that can also be applied to other hydride forming metals.   This up-to-date reference focuses on documented research surrounding DHC, including current methodologies for design and assessment of the results of periodic in-service inspections of pressure tubes in nuclear reactors. Emphasis is placed on showing that our understanding of DHC is supported by progress across a broad range of fields. These include hysteresis associated with first-order phase transformations; phase relationships in coherent crystalline metallic...

Considerations concerning the strategy of corium retention in the reactor vessel

International Nuclear Information System (INIS)

2015-01-01

Third-generation nuclear reactors are characterised by consideration during design of core meltdown accidents. More specifically, dedicated measures or devices must be implemented to avoid basemat melt-through in the reactor building. These devices must have a high level of confidence. The strategy of corium retention in the reactor vessel, if supported by appropriate research and development, makes it possible to achieve this objective. IRSN works alone or in partnerships to address all the issues associated with in-vessel corium retention. This document describes the in-vessel corium retention strategy and its limitations, along with the research programs conducted by IRSN in this area

Reactor Pressure Vessel Steels

Energy Technology Data Exchange (ETDEWEB)

Van de Velde, J.; Fabry, A.; Van Walle, E.; Chaoudi, R

1998-07-01

SCK-CEN's R and D programme on Reactor Pressure Vessel (RPV) Steels in performed in support of the RVP integrity assessment. Its main objectives are: (1) to develop enhanced surveillance concepts by applying micromechanics and fracture-toughness tests to small specimens, and by performing damage modelling and microstructure characterization; (2) to demonstrate the applied methodology on a broad database; (3) to achieve regulatory acceptance and industrial use. Progress and achievements in 1999 are reported.

Effect of in-core instrumentation mounting location on external reactor vessel cooling

International Nuclear Information System (INIS)

Suh, Jungsoo; Ha, Huiun

2017-01-01

Highlights: • Numerical simulations were conducted for the evaluation of an IVR-ERVC application. • The ULPU-V experiment was simulated for the validation of numerical method. • The effect of ICI mounting location on an IVR-ERVC application was investigated. • TM-ICI is founded to be superior to BM-ICI for successful application of IVR-ERVC. - Abstract: The effect of in-core instrumentation (ICI) mounting location on the application of in-vessel corium retention through external reactor vessel cooling (IVR-ERVC), used to mitigate severe accidents in which the nuclear fuel inside the reactor vessel becomes molten, was investigated. Numerical simulations of the subcooled boiling flow within an advanced pressurized-water reactor (PWR) in IVR-ERVC applications were conducted for the cases of top-mounted ICI (TM-ICI) and bottom-mounted ICI (BM-ICI), using the commercially available computational fluid dynamics (CFD) software ANSYS-CFX. Shear stress transport (SST) and the RPI model were used for turbulence closure and subcooled flow boiling, respectively. To validate the numerical method for IVR applications, numerical simulations of ULPU-V experiments were also conducted. The BM-ICI reactor vessel was modeled using a simplified design of an advanced PWR with BM-ICI; the TM-ICI counterpart was modeled by removing the ICI parts from the original geometry. It was found that TM-ICI was superior to BM-ICI for successful application of IVR-ERVC. For the BM-ICI case, the flow field was complicated because of the existence of ICIs and a significant temperature gradient was observed near the ICI nozzles on the lower part of the reactor vessel, where the ICIs were attached. These observations suggest that the existence of ICI below the reactor vessel hinders reactor vessel cooling.

Reactor vessel head permanent shield

International Nuclear Information System (INIS)

Hankinson, M.F.; Leduc, R.J.; Richard, J.W.; Malandra, L.J.

1989-01-01

A nuclear reactor is described comprising: a nuclear reactor pressure vessel closure head; control rod drive mechanisms (CRDMs) disposed within the closure head so as to project vertically above the closure head; cooling air baffle means surrounding the control rod drive mechanisms for defining cooling air paths relative to the control rod drive mechanisms; means defined within the periphery of the closure head for accommodating fastening means for securing the closure head to its associated pressure vessel; lifting lugs fixedly secured to the closure head for facilitating lifting and lowering movements of the closure head relative to the pressure vessel; lift rods respectively operatively associated with the plurality of lifting lugs for transmitting load forces, developed during the lifting and lowering movements of the closure head, to the lifting lugs; upstanding radiation shield means interposed between the cooling air baffle means and the periphery of the enclosure head of shielding maintenance personnel operatively working upon the closure head fastening means from the effects of radiation which may emanate from the control rod drive mechanisms and the cooling air baffle means; and connecting systems respectively associated with each one of the lifting lugs and each one of the lifting rods for connecting each one of the lifting rods to a respective one of each one of the lifting lugs, and for simultaneously connecting a lower end portion of the upstanding radiation shield means to each one of the respective lifting lugs

A Study on the Radial Hydride Assisted Delayed Hydride Cracking of Zircaloy

Energy Technology Data Exchange (ETDEWEB)

Jeong, Jin-Ho; Lee, Ji-Min; Kim, Yong-Soo [Hanyang University, Seoul (Korea, Republic of)

2015-05-15

Extensive studies have been done on understanding of DHC(Delayed hydride cracking) phenomenon since several zirconium alloy pressure tubes failed in nuclear reactor in the 1970s. Recently, long-term dry storage strategy has been considered seriously in order to manage spent nuclear fuel in Korea and other countries around the world. Consequentially, many researches have been investigated the degradation mechanisms which will threaten the spent fuel integrity during dry storage and showed that hydrogen related phenomenon such as hydride reorientation and DHC are the critical factors. Especially, DHC is the direct cracking mechanism which can cause not only a through-wall defect but also a radiation leak to the environment. In addition, DHC can be enhanced by radial hydride as reported by Kim who demonstrate that radial hydrides clearly act as crack linkage path. This phenomenon is known as the radial hydride assisted DHC (RHA-DHC). Therefore, study on DHC is essential to ensure the safety of spent fuel. Finite element analysis will be carried out for the stress gradient evaluation around notch tip. A variation in thermal cycle which leads to change in hydrogen solid solution trajectory may be required. If the radial hydride precipitates at notch tip, we will investigate what conditions should be met. Ultimately, we will suggest the regulation criteria for long-term dry storage of spent nuclear fuel.

Morphology study on the depleted uranium as hydriding/dehydriding cycles

Energy Technology Data Exchange (ETDEWEB)

Chung, Dong-you, E-mail: dongyou@nfri.re.kr [National Fusion Research Institute, Daejeon (Korea, Republic of); Yun, Sei-Hun; Kang, Hyun-Goo; Chang, Min Ho; Oh, Yun Hee [National Fusion Research Institute, Daejeon (Korea, Republic of); Kang, Kweon Ho; Woo, Yoon Myung [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2016-11-01

Depleted Uranium (DU) is one of the strongest candidates as a getter material of hydrogen isotopes in the nuclear fusion reactor. In this work, small DU lump specimen with 99.8% purity was prepared for observation of morphology variation as hydriding/dehydriding cycles. Hydriding/dehydriding of DU was carried out more than 10 cycles for powder preparation. The pulverized DU specimen was safely handled in the glove box under Argon gas condition to minimize contact with oxygen and humidity. The morphology change according to hydriding/dehydriding cycles was observed by visual cell reactor, optical microscope and scanning electron microscope. The first hydriding of the small DU sample has progressed slowly with surface enlargement and volume expansion as time passes. After third hydriding/dehydriding cycles, most of DU was pulverized. The powder fineness of DU developed as hydriding/dehydriding cycle progresses. But the agglomerates of fine DU particles were observed. It was confirmed that the DU particles exist as porous agglomerates. And the particle agglomerate shows poor fluidity and even has the cohesive force.

Application of high strength steel to nuclear reactor containment vessel

International Nuclear Information System (INIS)

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

1976-01-01

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

Heat load imposed on reactor vessels during in-vessel retention of core melts

Energy Technology Data Exchange (ETDEWEB)

Kim, Su-Hyeon; Chung, Bum-Jin, E-mail: bjchung@khu.ac.kr

2016-11-15

Highlights: • Angular heat load on reactor vessel by natural convection of oxide pool was measured. • High Ra was achieved by using mass transfer experiments based on analogy concept. • Measured Nusselt numbers agreed reasonably with the other existing studies. • Three different types of volumetric heat sources were compared. • They didn’t affect the heat flux of the top plate but affected those of the reactor vessel. - Abstract: We measured the heat load imposed on reactor vessels by natural convection of the oxide pool in severe accidents. Based on the analogy between heat and mass transfer, mass transfer experiments were performed using a copper sulfate electroplating system. A modified Rayleigh number of the order 10{sup 14} was achieved in a small facility with a height of 0.1 m. Three different types of volumetric heat sources were compared and the average Nusselt number of the curved surface was 39% lower, whereas in the case of the top plate was 6% higher than in previous studies with a two-dimensional geometry due to the high Sc value of this study. Reliable experimental data on the angular heat flux ratios were reported compared to those of the BALI and SIGMA CP facilities in terms of fluctuations and consistency.

Structural failure analysis of reactor vessels due to molten core debris

International Nuclear Information System (INIS)

Pfeiffer, P.A.

1993-01-01

Maintaining structural integrity of the reactor vessel during a postulated core melt accident is an important safety consideration in the design of the vessel. This paper addresses the failure predictions of the vessel due to thermal and pressure loadings from the molten core debris depositing on the lower head of the vessel. Different loading combinations were considered based on a wet or dry cavity and pressurization of the vessel based on operating pressure or atmospheric (pipe break). The analyses considered both short term (minutes) and long term (days) failure modes. Short term failure modes include creep at elevated temperatures and plastic instabilities of the structure. Long term failure modes are caused by creep rupture that lead to plastic instability of the structure. The analyses predict the reactor vessel will remain intact after the core melt has deposited on the lower vessel head

The method of life extension for the High Flux Isotope Reactor vessel

International Nuclear Information System (INIS)

Chang, Shib-Jung.

1995-01-01

The state of the vessel steel embrittlement as a result of neutron irradiation can be measured by its increase in the nil ductility temperature (NDT). This temperature is sometimes referred to as the brittle-ductile transition temperature (DBT) for fracture. The life extension of the High Flux Isotope Reactor (HFIR) vessel is calculated by using the method of fracture mechanics. A hydrostatic pressure test (hydrotest) is performed in order to determine a safe vessel static pressure. It is then followed by using fracture mechanics to project the reactor life from the safe hydrostatic pressure. The life extension calculation provides the following information on the remaining life of the reactor as a function of the nil ductility temperature increase: the probability of vessel fracture due to hydrotest vs vessel life at several hydrotest pressures; the hydrotest time interval vs the uncertainty of the nil ductility temperature increase rate; and the hydrotest pressure vs the uncertainty of the nil ductility temperature increase rate. It is understood that the use of a complete range of uncertainties of the nil ductility temperature increase is equivalent to the entire range of radiation damage that can be experienced by the vessel steel. From the numerical values for the probabilities of the vessel fracture as a result of hydrotest, it is estimated that the reactor vessel life can be extended up to 50 EFPY (100 MW) with the minimum vessel operating temperature equal to 85 degree F

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

In-vessel Retention Strategy for High Power Reactors - K-INERI Final Report (includes SBLB Test Results for Task 3 on External Reactor Vessel Cooling (ERVC) Boiling Data and CHF Enhancement Correlations)

Energy Technology Data Exchange (ETDEWEB)

F. B. Cheung; J. Yang; M. B. Dizon; J. Rempe

2005-01-01

In-vessel retention (IVR) of core melt is a key severe accident management strategy adopted by some operating nuclear power plants and proposed for some advanced light water reactors (ALWRs). If there were inadequate cooling during a reactor accident, a significant amount of core material could become molten and relocate to the lower head of the reactor vessel, as happened in the Three Mile Island Unit 2 (TMI-2) accident. If it is possible to ensure that the vessel head remains intact so that relocated core materials are retained within the vessel, the enhanced safety associated with these plants can reduce concerns about containment failure and associated risk. For example, the enhanced safety of the Westinghouse Advanced 600 MWe PWR (AP600), which relied upon External Reactor Vessel Cooling (ERVC) for IVR, resulted in the U.S. Nuclear Regulatory Commission (US NRC) approving the design without requiring certain conventional features common to existing LWRs. However, it is not clear that currently proposed external reactor vessel cooling (ERVC) without additional enhancements could provide sufficient heat removal for higher-power reactors (up to 1500 MWe). Hence, a collaborative, three-year, U.S. - Korean International Nuclear Energy Research Initiative (INERI) project was completed in which the Idaho National Engineering and Environmental Laboratory (INEEL), Seoul National University (SNU), Pennsylvania State University (PSU), and the Korea Atomic Energy Research Institute (KAERI) investigated the performance of ERVC and an in-vessel core catcher (IVCC) to determine if IVR is feasible for reactors up to 1500 MWe.

Leak detection device for nuclear reactor pressure vessel

International Nuclear Information System (INIS)

Ikeda, Jun.

1988-01-01

Purpose: To test the leakage of a nuclear reactor pressure vessel during stopping for a short period of time with no change to the pressure vessel itself. Constitution: The device of the present invention comprises two O-rings disposed on the flange surface that connects a pressure vessel main body and an upper cover, a leak-off pipeway derived from the gap of the O-rings at the flange surface to the outside of the pressure vessel, a pressure detection means connected to the end of the pipeway, a humidity detection means disposed to the lead-off pipeway, a humidity detection means disposed to the lead-off pipeway, and gas supply means and gas suction means disposed each by way of a check valve to a side pipe branched from the pipeway. After stopping the operation of the nuclear reactor and pressurizing the pressure vessel by filling water, gases supplied to the gap between the O-rings at the flange surface by opening the check valve. In a case where water in the pressure vessel should leak to the flange surface, when gas suction is applied by properly opening the check valve, increase in the humidity due to the steams of leaked water diffused into the gas is detected to recognize the occurrence of leakage. (Kamimura, M.)

System for cooling the upper wall of a nuclear reactor vessel

International Nuclear Information System (INIS)

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

1974-01-01

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

Reactor core of FBR type reactor

International Nuclear Information System (INIS)

Hayashi, Hideyuki; Ichimiya, Masakazu.

1994-01-01

A reactor core is a homogeneous reactor core divided into two regions of an inner reactor core region at the center and an outer reactor core region surrounding the outside of the inner reactor core region. In this case, the inner reactor core region has a lower plutonium enrichment degree and less amount of neutron leakage in the radial direction, and the outer reactor core region has higher plutonium enrichment degree and greater amount of neutron leakage in the radial direction. Moderator materials containing hydrogen are added only to the inner reactor core fuels in the inner reactor core region. Pins loaded with the fuels with addition of the moderator materials are inserted at a ratio of from 3 to 10% of the total number of the fuel pins. The moderator materials containing hydrogen comprise zirconium hydride, titanium hydride, or calcium hydride. With such a constitution, fluctuation of the power distribution in the radial direction along with burning is suppressed. In addition, an absolute value of the Doppler coefficient can be increased, and a temperature coefficient of coolants can be reduced. (I.N.)

The technology development for surveillance test of reactor vessel materials

Energy Technology Data Exchange (ETDEWEB)

Chang, Kee Ok; Kim, Byoung Chul; Lee, Sam Lai; Choi, Sun Phil; Park, Day Young; Choi, Kwen Jai

1997-12-01

Benchmark test was performed in accordance with the requirement of US NRC Reg. Guide DG-1053 for Kori unit-1 in order to determine best-estimated fast neutron fluence irradiated into reactor vessel. Since the uncertainty of radiation analysis comes from the calculation error due to neutron cross-section data, reactor core geometrical dimension, core source, mesh density, angular expansion and convergence criteria, evaluation of calculational uncertainty due to analytical method was performed in accordance with the regulatory guide and the proof was performed for entire analysis by comparing the measurement value obtained by neutron dosimetry located in surveillance capsule. Best-estimated neutron fluence in reactor vessel was calculated by bias factor, neutron flux measurement value/calculational value, from reanalysis result from previous 1st through 4th surveillance testing and finally fluence prediction was performed for the end of reactor life and the entire period of plant life extension. Pressurized thermal shock analysis was performed in accordance with 10 CFR 50.61 using the result of neutron fluence analysis in order to predict the life of reactor vessel material and the criteria of safe operation for Kori unit 1 was reestablished. (author). 55 refs., 55 figs.

Manipulator for testing a top-opened reactor pressure vessel

International Nuclear Information System (INIS)

Bauer, R.; Kastl, H.

1991-01-01

The design is described of a manipulator to be inserted into the inside of reactor pressure vessels opened at the top. The main components of the manipulator include a fixed column protruding into the pressure vessel and a support which is slidable on the column and carries the bearing component for the measuring, testing, inspection and repair instruments. The device includes a driving equipment for the support as well as the power supply for the sets accommodated on the support, with the aim to reduce the failure rate of the manipulator as a whole, shorten the time necessary for its assembling and thus the time of staying in the reactor pressure vessel and, at the same time, make its maintenance and operation easier. (Z.S.). 13 figs

Guide to the periodic inspection of nuclear reactor steel pressure vessels

International Nuclear Information System (INIS)

1969-01-01

This Guide is intended to provide general information and guidance to reactor owners or operators, inspection authorities, certifying authorities or regulatory bodies who are responsible for establishing inspection procedures for specific reactors or reactor types, and for the preparation of national codes or standards. The recommendations of the Guide apply primarily to water-cooled steel reactor vessels which are at a sufficiently early stage of design so that recommendations to provide accessibility for inspection can be incorporated into the early stages of design and inspection planning. However, much of the contents of the Guide are also applicable in part to vessels for other reactor types, such as gas-cooled, pressure-tube, or liquid-metal-cooled reactors, and also to some existing water-cooled reactors and reactors which are in advanced stage of design or construction. 46 refs, figs, 1 tab

Design of High Temperature Reactor Vessel Using ANSYS Software

International Nuclear Information System (INIS)

Bandriyana; Kasmudin

2003-01-01

Design calculation and evaluation of material strength for high temperature reactor vessel based on the design of HTR-10 high temperature reactor vessel were carried out by using the ANSYS 5.4 software. ANSYS software was applied to calculate the combined load from thermal and pressure load. Evaluation of material strength was performed by calculate and determine the distribution of temperature, stress and strain in the thickness direction of vessel, and compared with its material strength for designed. The calculation was based on the inner wall temperature of vessel of 600 o C and the outer temperature of 500 and 600 o C. Result of calculation gave the maximum stress for outer temperature of 600 o C was 288 N/ mm 2 and strain of 0.000187. For outer temperature of 500 o C the maximum stress was 576 N/ mm 2 and strain of 0.003. Based on the analysis result, the material of steel SA 516-70 with limited stress for design of 308 N/ mm 2 can be used for vessel material with outer wall temperature of 600 o C

Process and apparatus for adjusting a new upper reactor internals in a reactor vessel of a PWR

International Nuclear Information System (INIS)

Frizot, A.; Cadaureille, G.; Lalere, C.; Machuron, J.Y.

1987-01-01

On the new upper reactor internals is mounted devices for alignment and clearances, before introducing in the reactor vessel. After introducing alignment and clearances are measured. Adjustment pieces are provided for optimum clearances and alignment and fixed after removal from vessel. Decontamination is made by using water jets prior to fitting recess parts [fr

Nuclear reactor support and seismic restraint with in-vessel core retention cooling features

Energy Technology Data Exchange (ETDEWEB)

Edwards, Tyler A.; Edwards, Michael J.

2018-01-23

A nuclear reactor including a lateral seismic restraint with a vertically oriented pin attached to the lower vessel head and a mating pin socket attached to the floor. Thermally insulating materials are disposed alongside the exterior surface of a lower portion of the reactor pressure vessel including at least the lower vessel head.

Nuclear reactor installation with outer shell enclosing a primary pressure vessel

International Nuclear Information System (INIS)

1975-01-01

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

Analysis of toroidal vacuum vessels for use in demonstration sized tokamak reactors

International Nuclear Information System (INIS)

Culbert, M.E.

1978-07-01

The vacuum vessel component of the tokamak fusion reactor is the subject of this study. The main objective of this paper was to provide guidance for the structural design of a thin wall externally pressurized toroidal vacuum vessel. The analyses are based on the available state-of-the-art analytical methods. The shortcomings of these analytical methods necessitated approximations and assumptions to be made throughout the study. A principal result of the study has been the identification of a viable vacuum vessel design for the Demonstration Tokamak Hybrid Reactor (DTHR) and The Next Step (TNS) Reactor

Characterisation of hydrides in a zirconium alloy, by EBSD

International Nuclear Information System (INIS)

Ubhi, H.S.; Larsen, K.

2012-01-01

Zirconium alloys are used in nuclear reactors owing to their low capture cross-section for thermal neutrons and good mechanical and corrosion properties. However, they do suffer from delayed hydrogen cracking (DHC) due to formation of hydride particles. This study shows how the electron back-scatter diffraction (EBSD) technique can be used to characterise hydrides and their orientation relationship with the matrix. Hydrided EB weld specimens were prepared by electro-polishing, characterised using Oxford instruments AZtecHKL EBSD apparatus and software attached to a FEG SEM. Hydrides were found to exist as fine intra granular plates and having the Blackburn orientation relationship, i.e. (0002)Zr//(111)hydride and (1120)Zr//(1-10)hydride. The hydrides were also found to contain sigma 3 boundaries as well as local misorientations. (author)

Thermal Behavior of the Reactor Vessel Penetration Under External Vessel Cooling During a Severe Accident

International Nuclear Information System (INIS)

Kang, Kyoung-Ho; Park, Rae-Joon; Kim, Jong-Tae; Min, Byung-Tae; Lee, Ki-Young; Kim, Sang-Baik

2004-01-01

Experimental and analytical studies on the thermal behavior of reactor vessel penetration have been performed under external vessel cooling during a severe accident in the Korean next-generation reactor APR1400. Two types of tests, SUS-EXT and SUS-DRY with and without external vessel cooling, respectively, have been performed using sustained heating by an induction heater. Three tests have been carried out varying the cooling conditions at the vessel outer surface in the SUS-EXT tests. The experimental results have been thermally estimated using the LILAC computer code. The experimental results indicate that the inner surface of the vessel was ablated by the 45-mm thickness in the SUS-DRY test. Despite the total ablation of the welding material, the penetration was not ejected outside the vessel, which could be attributed to the thermal expansion of the penetration. Unlike the SUS-DRY test, the thickness of the ablation was ∼15 to 20 mm at most, so the welding was preserved in the SUS-EXT tests. It is concluded from the experimental results that the external vessel cooling highly affected the ablation configuration and the thermal behaviors of the vessel and the penetration. An increase in coolant mass flow rate from 0.047 to 0.152 kg/s had effects on the thermal behavior of the lower head vessel and penetration in the SUS-EXT tests. The LILAC analytical results on temperature distribution and ablation depth in the lower head vessel and penetration were very similar to the experimental results

Cooling system for the connecting rings of a fast neutron reactor vessel

International Nuclear Information System (INIS)

Martin, J.-P.; Malaval, Claude

1974-01-01

A description is given of a cooling system for the vessel connecting rings of a fast neutron nuclear reactor, particularly of a main vessel containing the core of the reactor and a volume of liquid metal coolant at high temperature and a safety vessel around the main vessel, both vessels being suspended to a rigid upper slab kept at a lower temperature. It is mounted in the annular space between the two vessels and includes a neutral gas circuit set up between the wall of the main vessel to be cooled and that of the safety vessel itself cooled from outer. The neutral gas system comprises a plurality of ventilators fitted in holes made through the thickness of the upper slab and opening on to the space between the two vessels. It also includes two envelopes lining the walls of these vessels, establishing with them small section channels for the circulation of the neutral gas cooled against the safety vessel and heated against the main vessel [fr

Outlines of guidelines for the inspection and evaluation of reactor vessel internals

International Nuclear Information System (INIS)

Seki, Hiroaki; Kobayashi, Hiroyuki; Nakano, Morihito; Murai, Soutarou; Nomoto, Toshiharu

2014-01-01

'The guideline committee for the inspection and evaluation of Reactor Vessel Internals' of JANSI (Japan Nuclear Safety Institute) has been developing many guidelines based on principle which the conservative methodology, and covered both individual inspection method of reactor internals and application of repair methods for reactor internals. In this paper, some aspects of the JANSI-VIP-03 (Guidelines for the inspection and evaluation of Reactor Vessel Internals, revised Dec.2013) which is summary document of the committee activity, are introduced. (author)

Problems in manufacturing and transport of pressure vessels of integral reactors

International Nuclear Information System (INIS)

Kralovec, J.

1997-01-01

Integral water-cooled reactors are typical with eliminating large-diameter primary pipes and placing primary components, i.e. steam generators and pressurizers in reactor vessels. This arrangement leads to reactor pressure vessels of large dimensions: diameters, heights and thick walls and subsequently to great weights. Thus, even medium power units have pressure vessels which are on the very limit of present manufacturing capabilities. Principal manufacturing and inspection operations as well as pertinent equipment are concerned: welding, cladding, heat treatment, machining, shop-handling, non-destructive testing, hydraulic pressure tests etc. Tile transport of such a large and heavy component makes a problem which effects its design as well as the selection of the plant site. Railway, road and ship are possible ways of transport each of them having its advantages and limitations. Specific features and limits of the manufacture and transport of large pressure vessels are discussed in the paper. (author)

Assessment of the integrity of WWER type reactor pressure vessels

International Nuclear Information System (INIS)

Brumovsky, M.

1995-01-01

Procedures are given for the assessment of the residual lifetime of reactor pressure vessels with respect to a sudden failure, the lifetime of vessels with defects disclosed during in-service inspections, and the fatigue or corrosion-mechanical lifetime. Also outlined are the ways of assessing the effects of major degradation mechanisms, i.e. radiation embrittlement, thermal aging, and fatigue damage, including the use of calculated values and experimental examination, by means of surveillance specimens in particular. All results of assessment performed so far indicate that the life of reactor pressure vessels at the Dukovany, Jaslovske Bohunice, and Temelin nuclear power plants is well secured. 7 figs., 3 refs

Annealing the reactor vessel at the Palisades Plant

International Nuclear Information System (INIS)

Fenech, R.A.

1996-01-01

In the way of background, Palisades was licensed in 1967 and went commercial in 1971. Jumping to two years ago, we faced at that time three issues that challenged our ability to operate to end-of-license, which would be 2007 without any extensions. The three items were regulatory performance, economic performance, and reactor vessel embrittlement. We had not been operating the plant with the kind of conservative decisions and with the kind of safety margins that one is expected to operate a plant in the United States at this time. Our economic performance was not satisfactory in that our capacity factor was low and our costs high. In the area of reactor vessel embrittlement, our analysis showed that we would reach the NRC screening criteria for embrittlement in the year 2004. Over the last two years, we have made significant improvements in the first two areas. Our decision-making has changed. Our performance, especially over the last year and a half, has been excellent. In addition, we have gotten our capacity factors up and our costs under control. Clearly, sustained performance is what is going to carry the day but from what we can see and from where we are, we are in more of a maintenance-of-performance than in a turn-around situation. On the other hand, in the area of reactor vessel embrittlement, about a year and a half ago we had a bit of a setback. We had taken material from retired steam generators that had welds identical to the welds in our reactor vessel. When we analyzed the welds from our steam generators, we were given some surprises about the chemistry makeup. When we applied the new information to our analysis, we changed the date on which we would reach our screening criteria from 2004 to late 1999

Light-water-reactor pressure-vessel surveillance dosimetry using solid-state track recorders

International Nuclear Information System (INIS)

Ruddy, F.H.; Roberts, J.H.; Gold, R.; Preston, C.C.

1983-07-01

The accumulation of neutron dose by the pressure vessel of an operating nuclear power plant results in damage in the form of steel embrittlement. In order to ascertain the safe operating lifetime of the reactor pressure vessel, dosimetric measurements must be made to evaluate the neutron dose to the pressure vessel and relate this dose to the cumulative radiation damage. Advanced dosimetry techniques are being evaluated for surveillance of operating reactors. Solid-state track recorder (SSTR) techniques are included among these advanced dosimetry techniques. Described herein are low neutron fluence calibration and standardization measurements that are being carried out in pressure vessel mockup benchmark neutron fields in the USA, Belgium, and England. In addition, high fluence SSTR dosimetry capsules have been irradiated with metallurgical specimens in a pressure vessel mockup facility. The design and deployment of advances SSTR dosimetry capsules in operating power reactors are also described

Role of radiation embrittlement in reactor vessel integrity assessment

International Nuclear Information System (INIS)

Marston, T.U.; Chexal, V.K.; Wyckoff, M.

1982-01-01

Reactor vessel integrity calculations are complex. The effect of radiation embrittlement on vessel material properties is a very important aspect of any vessel integrity evaluation. The importance of realistic (based on surveillance capsule results) rather than conservative estimates of the material properties (based on regulatory curves) cannot be overestimated. It is also important to make realistic thermal hydraulic and system operations assumptions. In addition, use of actual flaw sizes from in-service inspections (versus hypothetical flaw size selection) will promote realism. Important research results exist that need to be incorporated into the regulatory process. The authors believe results from current research and development efforts will demonstrate that, with reasonable assumptions and best estimate calculations, the safety of even the older reactor vessels with high copper content welds can be assured over their design lifetimes without the need for major fixes. The utilities, through EPRI and the vendors, have dedicated a significant effort to solving the pressurized thermal shock problem

The 1500 MW fast breeder reactor the double envelope-vessel anchored in concrete

International Nuclear Information System (INIS)

Bolvin, M.

1981-01-01

This paper givers an account of EDF investigations to reduce the investment costs of the 1500 MW Fast Reactor (RNR 1500) without prejudice to the safety requirements. It deals with the double envelope-vessel, designed to minimize radiation consequences in the event of accidental leakage in the main vessel. In the Fast Reactors in operation (PHOENIX), under construction (CRYS-MALVILLE), and under project (NR 1500), the double envelope-steel vessel hangs down from the upper part of the reactor block, its weight being approximately 300 t. In the new design, the vessel is fixed into the concrete which supports the main vessel, by means of steel anchors. A thermal insulation isolates it from the main vessel. The installation of coils in the concrete, next to the lining, allows for water circulation to cool the concrete. (orig./GL)

Radiation resistance of concrete of nuclear reactor vessel

International Nuclear Information System (INIS)

Belyakov, V.V.; Denisov, A.V.; Korenevskij, V.V.; Muzalevskij, L.P.; Dubrovskij, V.B.; Ivanov, D.A.; Nazarov, I.L.; Sashin, N.L.

1992-01-01

Results of calculational-experimental determination of radiation resistance for concrete bases on limestone gravel and quartz sand, which are the most perspective materials for manufacturing prestressed concrete of the VG-400 reactor vessel are considered. Material samples under investigation were irradiated in the channels of the IBR-2 research reactor for the purpose of the calcultional result verification

Light-water reactor pressure vessel surveillance standards

International Nuclear Information System (INIS)

Anon.

1981-01-01

The master matrix standard describes a series of standard practices, guides, and methods for the prediction of neutron-induced changes in light-water reactor (LWR) pressure vessel steels throughout a pressure vessel's service life. Some of these are existing American Society for Testing and Materials (ASTM) standards, some are ASTM standards that have been modified, and some are newly proposed ASTM standards. The current (1) scope, (2) areas of application, (3) interrelationships, and (4) status and time table of development, improvement, validation, and calibration for a series of 16 ASTM standards are defined. The standard also includes a discussion of LWR pressure vessel surveillance - justification, requirements, and status of work

Positioning means for circumferentially locating inspection apparatus in a nuclear reactor vessel

International Nuclear Information System (INIS)

Burns, D.C.

1980-01-01

There is provided for a reactor vessel inspection device a support ring sized to accommodate the circular path defined by three or more guide studs extending upwardly from the vessel. The support ring has at least three movably mounted guide stud bushings which can be positionally adjusted to align each bushing with one of the studs. When engaged, the guide studs and bushings yield a coarse positioning of the inspection device relative to the reactor vessel. Also provided are three support legs which are clamped to the support ring and dimensioned to an appropriate length. Two of the support legs have shoes clamped thereto, configured to rest on an internal circumferential flange within the reactor vessel. The third support leg is provided with a specially adapted shoe configured to engage a locating element, the exact position of which is known, within the vessel to achieve fine positioning of the inspection device relative to the reactor vessel. The support ring is additionally provided with an annular key which runs longitudinally about its outer periphery. Clamping means utilized to secure the guide stud bushings and the support legs to the support ring are provided with keyways to insure automatic self-alignment when fully tightened. (auth)

Radiation embrittlement in pressure vessels of power reactors

International Nuclear Information System (INIS)

Kempf, Rodolfo; Fortis, Ana M.

2007-01-01

It is presented the project to study the effect of lead factors on the mechanical behavior of Reactor Pressure Vessel steels. It is described the facility designed to irradiate Charpy specimens with V notch of SA-508 type 3 steel at power reactor temperature, installed in the RA-1 reactor. The objective is to obtain the fracture behavior of irradiated specimens with different lead factors and to know their dependence with the diffusion of alloy elements. (author) [es

Probabilistic fracture mechanics analysis of reactor vessels with low upper-shelf fracture toughness

International Nuclear Information System (INIS)

Yoon, K.K.

1993-01-01

A class of submerged-arc welds used in fabricating early reactor vessels has relatively high copper contents. Studies have shown that when such vessels are irradiated, the copper contributes to lowering the Charpy upper-shelf energy level. To address this concern, 10CFR50, Appendix G requires a fracture mechanics analysis to demonstrate an adequate margin of safety for continued service. The B and W Owners Group (B and WOG) has been accumulating J-resistance fracture toughness data for these weld metals. Based on a mathematical model derived from this B and WOG data base, the first Appendix G analysis was performed. Another important issue affecting reactor vessel integrity is pressurized thermal shock (PIS) transients. In the early 1980s, probabilistic fracture mechanics analyses were performed on a reactor vessel to determine the probability of failure under postulated accident scenarios. Results of such analyses were used by the Nuclear Regulatory Commission (NRC) to establish the screening criteria for assessing reactor vessel integrity under PTS transient loads. This paper addresses the effect of low upper-shelf toughness on the probability of failure of reactor vessels under PTS loads. Probabilistic fracture mechanics codes were modified to include the low upper-shelf toughness model used in a reference and a series of analyses was performed using plant-specific material conditions and realistic PTS scenarios. The results indicate that low upper-shelf toughness has an insignificant effect on the probability of reactor vessel failures. This is mostly due to PTS transients being susceptible to crack initiation at low temperatures and not affected by upper-shelf fracture toughness

Research to sustain cases for Magnox-reactor steel pressure vessels

International Nuclear Information System (INIS)

Graham, W.J.

1997-01-01

Britain's Magnox Electric plc owns and operates six power stations, each of which has twin gas-cooled reactors of the Magnox-fuel type. The older group of four power stations has steel pressure-circuits. The reactor cores are housed within spherical, steel vessels. This article describes some of the research which is undertaken to sustain the safety cases for these steel vessels which have now been in operation for just over 30 years. (author) 2 figs., 4 refs

Hydride blister formation simulation in Candu type reactors

International Nuclear Information System (INIS)

Otero, D.; Bollini, C.; Sangregorio, D.

1992-01-01

We have developed a computer code for the probability study of hydride blister formation in pressure tubes named BLIFO. The basic hypothesis of the model are: the pressure tube is divided into five areas according to the existence of four garter springs. For each area the probability of blister formation is the probability of the hydrogen content exceeding a critical threshold when contact tube is present; the probability of a blister in a tube is the OR combination of the probabilities of a blister in each area; the tube contact is a function of the garter springs location, and the time; the critical hydrogen threshold is sorted over the areas within the pressure tube; hydrogen pick-up rate was sorted with a Gaussian distribution; the initial hydrogen content values for each tube were measured before the ensamble and they are used in the code. For Embalse evaluation, we build up a subroutine that simulate Gaussian distribution using the parameters of a typical nuclear power Candu reactor garter spring distribution. (author)

Design and analysis of prestressed reactor vessels

International Nuclear Information System (INIS)

Burrow, R.E.D.

1978-01-01

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

TMI-2 reactor-vessel head removal and damaged-core-removal planning

International Nuclear Information System (INIS)

Logan, J.A.; Hultman, C.W.; Lewis, T.J.

1982-01-01

A major milestone in the cleanup and recovery effort at TMI-2 will be the removal of the reactor vessel closure head, planum, and damaged core fuel material. The data collected during these operations will provide the nuclear power industry with valuable information on the effects of high-temperature-dissociated coolant on fuel cladding, fuel materials, fuel support structural materials, neutron absorber material, and other materials used in reactor structural support components and drive mechanisms. In addition, examination of these materials will also be used to determine accident time-temperature histories in various regions of the core. Procedures for removing the reactor vessel head and reactor core are presented

Prestressed concrete vessels suitable for helium high temperature reactors

International Nuclear Information System (INIS)

Lockett, G.E.; Kinkead, A.N.

1967-02-01

In considering prestressed concrete vessels for use with helium cooled high temperature reactors, a number of new problems arise and projected designs involve new approaches and new solutions. These reactors, having high coolant outlet temperature from the core and relatively high power densities, can be built into compact designs which permit usefully high working pressures. Consequently, steam generators and circulating units tend to be small. Although circuit activity can be kept quite low with coated particle fuels, designs which involve entry for subsequent repair are not favoured, and coupled with the preferred aim of using fully shop fabricated units within the designs with removable steam generators which involve no tube welding inside the vessel. A particular solution uses a number of slim cylindrical assemblies housed in the wall of the pressure vessel and this vessel design concept is presented. The use of helium requires very high sealing standards and one of the important requirements is a vessel design which permits leak testing during construction, so that a repair seal can be made to any faulty part in a liner seam. Very good demountable joint seals can be made without particular difficulty and Dragon experience is used to provide solutions which are suitable for prestressed concrete vessel penetrations. The concept layout is given of a vessel meeting these requirements; the basis of design is outlined and special features of importance discussed. (author)

Microstructural evolution in neutron irradiated reactor pressure vessel steels

International Nuclear Information System (INIS)

English, C.A.; Phythian, W.J.

1998-01-01

As a result of the popularity of the Agencies report 'Neutron Irradiation Embrittlement of Reactor Pressure Vessel Steels' of 1975, it was decided that another report on this broad subject would be of use. In this report, background and contemporary views on specially identified areas of the subject are considered as self-contained chapters, written by experts. The microstructural evolution in neutron irradiated reactor pressure vessel steels is described. The damage mechanisms are elaborated and techniques for examining the microstructure are suggested. The importance of the initial damage event is analysed, and the microstructural evolution in RPV steels is examined

Prestressed concrete pressure vessels for boiling water reactors

International Nuclear Information System (INIS)

Menon, S.

1979-12-01

Following a general description of the Scandinavian cooperative project on prestressed concrete pressure vessels for boiling water reactors, detailed discussion is given in four appendices of the following aspects: the verification programme of tests and studies, the development and testing of a liner venting system, a preliminary safety philosophy and comparative assessment of cold and hot liners. Vessel failure probability is briefly discussed and some figures presented. The pressure gradients in the vessel wall resulting from various stipulated linear cracks, with a liner venting system are presented graphically. (JIW)

Multidimensional simulations of hydrides during fuel rod lifecycle

International Nuclear Information System (INIS)

Stafford, D.S.

2015-01-01

In light water reactor fuel rods, waterside corrosion of zirconium-alloy cladding introduces hydrogen into the cladding, where it is slightly soluble. When the solubility limit is reached, the hydrogen precipitates into crystals of zirconium hydride which decrease the ductility of the cladding and may lead to cladding failure during dry storage or transportation events. The distribution of the hydride phase and the orientation of the crystals depend on the history of the spatial temperature and stress profiles in the cladding. In this work, we have expanded the existing hydride modeling capability in the BISON fuel performance code with the goal of predicting both global and local effects on the radial, azimuthal and axial distribution of the hydride phase. We compare results from 1D simulations to published experimental data. We demonstrate the new capability by simulating in 2D a fuel rod throughout a lifecycle that includes irradiation, short-term storage in the spent fuel pool, drying, and interim storage in a dry cask. Using the 2D simulations, we present qualitative predictions of the effects of the inter-pellet gap and the drying conditions on the growth of a hydride rim. - Highlights: • We extend BISON fuel performance code to simulate lifecycle of fuel rods. • We model hydrogen evolution in cladding from reactor through dry storage. • We validate 1D simulations of hydrogen evolution against experiments. • We show results of 2D axisymmetric simulations predicting hydride formation. • We show how our model predicts formation of a hydride rim in the cladding.

Prestressed concrete reactor vessel for the HHT-670 MW(e) demonstration plant. Pt.1. Design of the multi-cavity prestressed concrete reactor vessel with warm liner

International Nuclear Information System (INIS)

Lafitte, R.; Marchand, J.D.

1979-01-01

The design studies and tests described in this paper were undertaken as part of ''PROJECT HHT'', a German-Swiss joint effort for the development of high-temperature helium cooled reactors with direct-cycle turbine. The prestressed concrete reactor pressure vessel encloses the core of the reactor itself, the heat exchangers (coolers and recuperators), the helium turbine, the main helium circuit, all nuclear and thermal equipment, and auxiliary reactor cooling equipment. In order to make the liner accessible for inspection, no thermal insulation is provided between the coolant and the liner. The temperature of the helium in contact with the liner is limited to 200 0 C, under all normal operation conditions of the reactor. In the HHT reactor pressure vessel, the resisting structure is protected thermally by a layer of warm concrete between the liner and the structural prestressed concrete. The main features of this pressure vessel are the marked pressure differences in the cavities during normal operation, and the use of warm liner. The objectives of the reference design were chiefly related to the sizing up of the main structure, taking into account the modifications to be expected in the material characteristics as a result of the high temperatures developed

Investigation of the failure of a reactor pressure vessel by plastic instability

International Nuclear Information System (INIS)

Laemmer, H.; Ritter, B.

1994-01-01

A possible consequence of a core meltdown accident in a pressurized water reactor is the failure of the reactor pressure vessel under high internal pressure. With the aid of the finite element program ABAQUS and using a material model of the thermo-plasticity for large deformation, the failure of the reactor pressure vessel due to plastic instability was examined. It was apparent from the finite element calculations that solely due to reduction in strength of the material, even for internal wall temperatures clearly below the core melt; of about 2000 C, the critical internal pressure can fall to values which are lower than the working pressure. With the aid of simplified geometry, a lower limit for the pressure at failure of the reactor pressure vessel can be calculated. (orig./HP) [de

Basic conceptions for reactor pressure vessel manipulators and their evaluation

International Nuclear Information System (INIS)

Popp, P.

1987-01-01

The study deals with application fields and basic design conceptions of manipulators in reactor pressure vessels as well as their evaluation. It is shown that manipulators supported at the reactor flange have essential advantages

Standard Practice for Design of Surveillance Programs for Light-Water Moderated Nuclear Power Reactor Vessels

CERN Document Server

American Society for Testing and Materials. Philadelphia

2010-01-01

1.1 This practice covers procedures for designing a surveillance program for monitoring the radiation-induced changes in the mechanical properties of ferritic materials in light-water moderated nuclear power reactor vessels. This practice includes the minimum requirements for the design of a surveillance program, selection of vessel material to be included, and the initial schedule for evaluation of materials. 1.2 This practice was developed for all light-water moderated nuclear power reactor vessels for which the predicted maximum fast neutron fluence (E > 1 MeV) at the end of license (EOL) exceeds 1 × 1021 neutrons/m2 (1 × 1017 n/cm2) at the inside surface of the reactor vessel. 1.3 This practice applies only to the planning and design of surveillance programs for reactor vessels designed and built after the effective date of this practice. Previous versions of Practice E185 apply to earlier reactor vessels. 1.4 This practice does not provide specific procedures for monitoring the radiation induced cha...

Nuclear reactor having an inflatable vessel closure seal structure

International Nuclear Information System (INIS)

1980-01-01

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

Assessment of reactor vessel integrity (ARVI)

Energy Technology Data Exchange (ETDEWEB)

Sehgal, B.R. E-mail: sehgal@ne.kth.se; Theerthan, A.; Giri, A.; Karbojian, A.; Willschuetz, H.G.; Kymaelaeinen, O.; Vandroux, S.; Bonnet, J.M.; Seiler, J.M.; Ikkonen, K.; Sairanen, R.; Bhandari, S.; Buerger, M.; Buck, M.; Widmann, W.; Dienstbier, J.; Techy, Z.; Kostka, P.; Taubner, R.; Theofanous, T.; Dinh, T.N

2003-04-01

The cost-shared project ARVI (assessment of reactor vessel integrity) involves a total of nine organisations from Europe and USA. The objective of the ARVI Project is to resolve the safety issues that remain unresolved for the melt vessel interaction phase of the in-vessel progression of a severe accident. The work consists of experiments and analysis development. Four tests were performed in the EC-FOREVER Programme, in which failure was achieved in-vessels employing the French pressure vessel steel. The tests were analysed with the commercial code ANSYS-Multiphysics, and the codes SYSTUS+ and PASULA, and quite good agreement was achieved for the failure location. Natural convection experiments in stratified pools have been performed in the SIMECO and the COPO facilities, which showed that much greater heat is transferred downwards for immiscible layers or before layers mix. A model for gap cooling and a set of simplified models for the system codes have been developed. MVITA code calculations have been performed for the Czech and Hungarian VVERs, towards evaluation of the in-vessel melt retention accident management scheme. Tests have been performed at the ULPU facility with organised flow for vessel external cooling. Considerable enhancement of the critical heat flux (CHF) was obtained. The ARVI Project has reached the halfway stage. This paper presents the results obtained thus far from the project.

Correlation between radiation damage and magnetic properties in reactor vessel steels

Energy Technology Data Exchange (ETDEWEB)

Kempf, R.A., E-mail: kempf@cnea.gov.ar [División Caracterización, GCCN, CAC-CNEA (Argentina); Sacanell, J. [Departamento Física de la Materia Condensada, GIyA, CAC-CNEA, CONICET (Argentina); Milano, J. [División Resonancias Magnéticas, CAB-CNEA, CONICET (Argentina); Guerra Méndez, N. [Departamento Física de la Materia Condensada, GIyA, CAC-CNEA, CONICET (Argentina); Winkler, E.; Butera, A. [División Resonancias Magnéticas, CAB-CNEA, CONICET (Argentina); Troiani, H. [División Física de Metales, CAB-CNEA and Instituto Balseiro (UNCU), CONICET (Argentina); Saleta, M.E. [División Resonancias Magnéticas, CAB-CNEA, CONICET (Argentina); Fortis, A.M. [Departamento Estructura y Comportamiento. Gerencia Materiales-GAEN, CAC-CNEA (Argentina)

2014-02-01

Since reactor pressure vessel steels are ferromagnetic, provide a convenient means to monitor changes in the mechanical properties of the material upon irradiation with high energy particles, by measuring their magnetic properties. Here, we discuss the correlation between mechanical and magnetic properties and microstructure, by studying the flux effect on the nuclear pressure vessel steel used in reactors currently under construction in Argentina. Charpy-V notched specimens of this steel were irradiated in the RA1 experimental reactor at 275 °C with two lead factors (LFs), 93 and 183. The magnetic properties were studied by means of DC magnetometry and ferromagnetic resonance. The results show that the coercive field and magnetic anisotropy spatial distribution are sensitive to the LF and can be explained by taking into account the evolution of the microstructure with this parameter. The saturation magnetization shows a dominant dependence on the accumulated damage. Consequently, the mentioned techniques are suitable to estimate the degradation of the reactor vessel steel.

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

Energy Technology Data Exchange (ETDEWEB)

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

2001-03-15

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

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

International Nuclear Information System (INIS)

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

2001-03-01

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

Experimental Investigation of Creep Behavior of Reactor Vessel Lower Head

International Nuclear Information System (INIS)

Chu, T.Y.; Pilch, M.; Bentz, J.H.; Behbahani, A.

1999-01-01

The authors report a study which aimed at experimentally and numerically investigating and characterizing the failure of a reactor pressure vessel (RPV) lower head due to thermal and pressure loads generated by a severe accident. They present the experimental apparatus which is based on a scaled version of the lower part of a TMI-like reactor pressure vessel without vessel skirt. They report and comment the results obtained during the first five experiments: uniform heating and non penetrations, centre-peaked heat flux and no penetrations, edge-peaked heat flux and no penetrations, uniform heating with penetrations, edge-peaked heat flux with penetrations. They compare the third and fifth experience (those with edge-peaked heat flux)

Completely integrated prestressed-concrete reactor pressure vessel, type 'Star'

International Nuclear Information System (INIS)

Neunert, B.; Jueptner, G.; Kumpf, H.

1975-01-01

The star support vessel is suitable for the connection to all primary circuit systems consisting of a main vessel and a number of satellite vessels around and connected to it, i.e. for LWR, HTR and process reactor. It must be made clear, however, that the PWR in particular with its components does not appear to be suited for the optimum incorporation in a prestressed-concrete pressure vessel system, no matter what kind. There are clear concepts about modifications which, however, require considerable development expenditure. (orig./LH) [de

Coupled thermo-mechanical creep analysis for boiling water reactor pressure vessel lower head

International Nuclear Information System (INIS)

Villanueva, Walter; Tran, Chi-Thanh; Kudinov, Pavel

2012-01-01

Highlights: ► We consider a severe accident in a BWR with melt pool formation in the lower head. ► We study the influence of pool depth on vessel failure mode with creep analysis. ► There are two modes of failure; ballooning of vessel bottom and a localized creep. ► External vessel cooling can suppress creep and subsequently prevent vessel failure. - Abstract: In this paper we consider a hypothetical severe accident in a Nordic-type boiling water reactor (BWR) at the stage of relocation of molten core materials to the lower head and subsequent debris bed and then melt pool formation. Nordic BWRs rely on reactor cavity flooding as a means for ex-vessel melt coolability and ultimate termination of the accident progression. However, different modes of vessel failure may result in different regimes of melt release from the vessel, which determine initial conditions for melt coolant interaction and eventually coolability of the debris bed. The goal of this study is to define if retention of decay-heated melt inside the reactor pressure vessel is possible and investigate modes of the vessel wall failure otherwise. The mode of failure is contingent upon the ultimate mechanical strength of the vessel structures under given mechanical and thermal loads and applied cooling measures. The influence of pool depth and respective transient thermal loads on the reactor vessel failure mode is studied with coupled thermo-mechanical creep analysis. Efficacy of control rod guide tube (CRGT) cooling and external vessel wall cooling as potential severe accident management measures is investigated. First, only CRGT cooling is considered in simulations revealing two different modes of vessel failure: (i) a ‘ballooning’ of the vessel bottom and (ii) a ‘localized creep’ concentrated within the vicinity of the top surface of the melt pool. Second, possibility of in-vessel retention with CRGT and external vessel cooling is investigated. We found that the external vessel

Evaluation of HFIR [High Flux Isotope Reactor] pressure-vessel integrity considering radiation embrittlement

International Nuclear Information System (INIS)

Cheverton, R.D.; Merkle, J.G.; Nanstad, R.K.

1988-04-01

The High Flux Isotope Reactor (HFIR) pressure vessel has been in service for 20 years, and during this time, radiation damage was monitored with a vessel-material surveillance program. In mid-November 1986, data from this program indicated that the radiation-induced reduction in fracture toughness was greater than expected. As a result, a reevaluation of vessel integrity was undertaken. Updated methods of fracture-mechanics analysis were applied, and an accelerated irradiations program was conducted using the Oak Ridge Research Reactor. Results of these efforts indicate that (1) the vessel life can be extended 10 years if the reactor power level is reduced 15% and if the vessel is subjected to a hydrostatic proof test each year; (2) during the 10-year life extension, significant radiation damage will be limited to a rather small area around the beam tubes; and (3) the greater-than-expected damage rate is the result of the very low neutron flux in the HFIR vessel relative to that in samples of material irradiated in materials-testing reactors (a factor of ∼10 4 less), that is, a rate effect

Life extension of the St. Lucie unit 1 reactor vessel

International Nuclear Information System (INIS)

Rowan, G.A.; Sun, J.B.; Mott, S.L.

1991-01-01

In late 1989, Florida Power and Light Company (FP and L) established the policy that St. Lucie unit 1 should not be prevented from achieving a 60-yr operating life by reactor vessel embrittlement. A 60-yr operating life means that the plant would be allowed to operate until the year 2036, which is 20 years beyond the current license expiration date of 2016. Since modifications to the reactor vessel and its components are projected to be expensive, the desire of FP and L management was to achieve this lifetime extension through the use of fuel management and proven technology. The following limitations were placed on any acceptable method for achieving this lifetime extension capability: low fuel cycle cost; low impact on safety parameters; very little or no operations impact; and use of normal reactor materials. A task team was formed along with the Advanced Nuclear Fuels Company (ANF) to develop a vessel-life extension program

Pressurized water reactor with a reactor pressure vessel

International Nuclear Information System (INIS)

Werres, L.

1979-01-01

The core barrel is suspended from a flange by means of a grid. The coolant enters the barrel from below through the grid. In order to get a uniform flow over the reactor core there is provided for a guiding device below the grid. It consists of a cylindrical shell with borings uniformly distributed around the shell as well as fins on the inner surface of the shell and slots at the bottom facing the pressure vessel. (GL) [de

To the problem of reinforced concrete reactor vessel design and calculation

International Nuclear Information System (INIS)

Kirillov, A.P.; Artem'ev, V.P.; Bogopol'skij, V.G.; Nikolaev, Yu.B.; Paushkin, A.G.

1980-01-01

Modern methods for calculating reactor vessels of prestressed reinforced concrete are analyzed. It is shown that during the stage of technical and economical substantiation of reactor vessel structure for determining its stressed-deformed state engineering methods of calculation must be used, in particular, fragmentation method, method of rings and plates, and during the stages of contract and detail designs - method of finite elements and dynamic relaxation method. It is concluded that when solving cyclic symmetrical problems as well as asymmetrical problems, calculational algorithms for axis-symmetrical distributions of stresses in the vessel with provision for elastic properties of structural material may be used

Electroerosion cutting of low-sized templets from WWER-1000 type reactor vessel

International Nuclear Information System (INIS)

Neklyudov, I.M.; Ozhigov, L.S.; Gozhenko, S.V.

2012-01-01

The article presents the results of developed method of electroerosion cutting of low-sized templets for the reactor vessel metal composition and structure control in laboratory environment. The article describes the equipment for the remote electroerosive cutting of templets from WWER-1000 type reactor vessel by rigid electrode. The testing results are also shown.

Measuring neutron flux density in near-vessel space of a commercial WWER-1000 reactor

International Nuclear Information System (INIS)

Borodkin, G.I.; Eremin, A.N.; Lomakin, S.S.; Morozov, A.G.

1987-01-01

Distribution of neutron flux density in two experimental channels on the reactor vessel external surface and in ionization chamber channel of a commercial WWER-1000 reactor, is measured by the activation detector technique. Azimuthal distributions of fast and thermal neutron fluxes and height distributions of fast neutron flux density within energy range >1.2 and 2.3 MeV are obtained. Conclusion is made, that reactor core state and its structural peculiarities in the measurement range essentially affect space and energy distribution of neutron field near the vessel. It should be taken into account when determining permissible neutron fluence for the reactor vessel

Adoption of in-vessel retention concept for VVER-440/V213 reactors in Central European Countries

Energy Technology Data Exchange (ETDEWEB)

Matejovic, Peter, E-mail: peter.matejovic@ivstt.sk [Inzinierska Vypoctova Spolocnost (IVS), Jana Holleho 5, 91701 Trnava (Slovakia); Barnak, Miroslav; Bachraty, Milan; Vranka, Lubomir [Inzinierska Vypoctova Spolocnost (IVS), Jana Holleho 5, 91701 Trnava (Slovakia); Berky, Robert [Integrita a Bezpecnost Ocelovych Konstrukcii, Rybnicna 40, 831 07 Bratislava (Slovakia)

2017-04-01

Highlights: • Design of in-vessel retention concept for VVER-440/V213 reactors. • Thermal loads acting on the inner reactor surface. • Structural response of reactor pressure vessel. • External reactor vessel cooling. - Abstract: An in-vessel retention (IVR) concept was proposed for standard VVER-440/V213 reactors equipped with confinement made of reinforced concrete and bubbler condenser pressure suppression system. This IVR concept is based on simple modifications of existing plant technology and thus it was attractive for plant operators in Central European Countries. Contrary to the solution that was adopted before at Loviisa NPP in Finland (two units of VVER-440/V213 reactor with steel confinement equipped with ice condenser), the coolant access to the reactor pressure vessel from flooded cavity is enabled via closable hole installed in the centre of thermal shield of the reactor lower head instead of lowering this massive structure in the case of severe accident. As a consequence, the crucial point of this IVR concept is narrow gap between torispherical lower head and thermal and biological shield. Here the highest thermal flux is expected in the case of severe accident. Thus, realistic estimation of thermal load and corresponding deformations of reactor wall and their impact on gap width for coolant flow are of primarily importance. In this contribution the attention is paid especially to the analytical support with emphasis to the following points: 1) {sup ∗}Estimation of thermal loads acting on the inner reactor surface; 2) {sup ∗}Estimation of structural response of reactor pressure vessel (RPV) with emphasis on the deformation of outer reactor surface and its impact on the annular gap between RPV wall and thermal/biological shield; 3) {sup ∗}Analysis of external reactor vessel cooling. For this purpose the ASTEC code was used for performing analysis of core degradation scenarios, the ANSYS code for structural analysis of reactor vessel

The characteristics of the prestressed concrete reactor vessel of the HHT demonstration plant

International Nuclear Information System (INIS)

Schoening, J.; Schwiers, H.G.

1979-01-01

The paper concentrates on the design studies of the HTGR prestressed concrete reactor vessel (PCRV) for the HHT Demonstration Plant. The multi-cavity reactor pressure vessel accommodates all components carrying primary gas, including heat exchangers and gas turbine. For reasons of economics and availability of the reactor plant, generic requirements are made for the PCRV. A short description of the power plant is also presented

Determination of hydrogen in zirconium hydride and uranium-zirconium hydride by inert gas exraction-gravimetric method

International Nuclear Information System (INIS)

Hoshino, Akira; Iso, Shuichi

1976-01-01

An inert gas extraction-gravimetric method has been applied to the determination of hydrogen in zirconium hydride and uranium-zirconium hydride which are used as neutron moderator and fuel of nuclear safety research reactor (NSRR), respectively. The sample in a graphite-enclosed quartz crucible is heated inductively to 1200 0 C for 20 min in a helium stream. Hydrogen liberated from the sample is oxidized to water by copper(I) oxide-copper(II) oxide at 400 0 C, and the water is determined gravimetrically by absorption in anhydrone. The extraction curves of hydrogen for zirconium hydride and uranium-zirconium hydride samples are shown in Figs. 2 and 3. Hydrogen in the samples is extracted quantitatively by heating at (1000 -- 1250) 0 C for (10 -- 40) min. Recoveries of hydrogen in the case of zirconium hydride were examined as follows: a weighed zirconium rod (5 phi x 6 mm, hydrogen -5 Torr. After the chamber was filled with purified hydrogen to 200 Torr, the rod was heated to 400 0 C for 15 h, and again weighed to determine the increase in weight. Hydrogen in the rod was then determined by the proposed method. The results are in excellent agreement with the increase in weight as shown in Table 1. Analytical results of hydrogen in zirconium hydride samples and an uranium-zirconium hydride sample are shown in Table 2. (auth.)

Image processing algorithm for robot tracking in reactor vessel

International Nuclear Information System (INIS)

Kim, Tae Won; Choi, Young Soo; Lee, Sung Uk; Jeong, Kyung Min; Kim, Nam Kyun

2011-01-01

In this paper, we proposed an image processing algorithm to find the position of an underwater robot in the reactor vessel. Proposed algorithm is composed of Modified SURF(Speeded Up Robust Feature) based on Mean-Shift and CAMSHIFT(Continuously Adaptive Mean Shift Algorithm) based on color tracking algorithm. Noise filtering using luminosity blend method and color clipping are preprocessed. Initial tracking area for the CAMSHIFT is determined by using modified SURF. And then extracting the contour and corner points in the area of target tracked by CAMSHIFT method. Experiments are performed at the reactor vessel mockup and verified to use in the control of robot by visual tracking

Rupture tests with reactor pressure vessel head models

International Nuclear Information System (INIS)

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

2003-01-01

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

A structure for the protection of nuclear-reactor pressurized-vessels against rupture

International Nuclear Information System (INIS)

Marcellin, J.-P.; Aubert, Gilles

1974-01-01

Description is given of a structure for the protection of nuclear-reactor pressurized-vessels against rupture. Said structure comprises a pre-stressed concrete tank adapted to surround the tank side-wall and bottom, said tank being higher than said vessel, said tank being provided with ports for passing cooling fluid ducts therethrough, and a crown adapted to rest along the periphery of the reactor-cover and made integral therewith. This can be applied to reactors of the PWR type [fr

A model for structural analysis of nuclear reactor pressure vessel flanges

International Nuclear Information System (INIS)

Oliveira, C.A. de.

1987-01-01

Due to the recent Brazilian advances in the nuclear technology area, it has been necessary the development of design and analysis methods for pressurized water reactor components, also as other components of a nuclear plant. This work proposes a methodology for the structural analysis of large diameter nuclear reactor pressure vessel flanges. In the analysis the vessel is divided into shell-of-revolution elements, the flanges are represented by rigid rings, and the bolts are treated as beams. The flexibility method is used for solving the problem. A computer program is shown, and the given results (displacements and stresses) are compared with results obtained by the finite element method. Although developed for nuclear reactor pressure vessel calculations, the program is more general, being possible its use for the analysis of any structure composed by shells of revolution. (author)

Evaluation of the integrity of SEP reactor vessels

International Nuclear Information System (INIS)

Hoge, K.G.

1979-12-01

A documented review is presented of the integrity of the 11 reactor pressure vessels covered in the Systematic Evaluation Program. This review deals primarily with the design specifications and quality assurance programs used in the vessel construction and the status of material surveillance programs, pressure-temperature operating limits, and inservice inspection programs of the applicable plants. Several generic items such as PWR overpressurization protection and BWR nozzle and safe-end cracking also are evaluated. The 11 vessels evaluated include Dresden Units 1 and 2, Big Rock Point, Haddam Neck, Yankee Rowe, Oyster Creek, San Onofre 1, LaCrosse, Ginna, Millstone 1, and Palisades

Characteristics of hydride precipitation and reorientation in spent-fuel cladding

International Nuclear Information System (INIS)

Chung, H. M.; Strain, R. V.; Billone, M. C.

2000-01-01

The morphology, number density, orientation, distribution, and crystallographic aspects of Zr hydrides in Zircaloy fuel cladding play important roles in fuel performance during all phases before and after discharge from the reactor, i.e., during normal operation, transient and accident situations in the reactor, temporary storage in a dry cask, and permanent storage in a waste repository. In the past, partly because of experimental difficulties, hydriding behavior in irradiated fuel cladding has been investigated mostly by optical microscopy (OM). In the present study, fundamental metallurgical and crystallographic characteristics of hydride precipitation and reorientation were investigated on the microscopic level by combined techniques of OM and transmission electron and scanning electron microscopy (TEM and SEM) of spent-fuel claddings discharged from several boiling and pressurized water reactors (BWRs and PWRs). Defueled sections of standard and Zr-lined Zircaloy-2 fuel claddings, irradiated to fluences of ∼3.3 x 10 21 n cm -2 and ∼9.2 x 10 21 n cm -2 (E > 1 MeV), respectively, were obtained from spent fuel rods discharged from two BWRs. Sections of standard and low-tin Zircaloy-4 claddings, irradiated to fluences of ∼4.4 x 10 21 n cm -2 , ∼5.9 x 10 21 n cm -2 , and ∼9.6 x 10 21 n cm -2 (E > 1 MeV) in three PWRs, were also obtained. Microstructural characteristics of hydrides were analyzed in as-irradiated condition and after gas-pressurization-burst or expanding-mandrel tests at 292-325 C in Ar for some of the spent-fuel claddings. Analyses were also conducted of hydride habit plane, morphology, and reorientation characteristics on unirradiated Zircaloy-4 cladding that contained dense radial hydrides. Reoriented hydrides in the slowly cooled unirradiated cladding were produced by expanding-mandrel loading

An evaluation of alternative reactor vessel cutting technologies for the decommissioning of the experimental boiling water reactor at Argonne National Laboratory

International Nuclear Information System (INIS)

Boing, L.E.; Henley, D.R.; Manion, W.J.; Gordon, J.W.

1991-01-01

This paper will detail (1) a brief overview of the current status of the EBWR D ampersand D Project, and (2) the results of a study performed to evaluate the metal cutting technologies available to size reduce the EBWR reactor vessel. The techniques evaluated were: Plasma arc, Arc saw, Oxyacetylene, Electric arc gouging, Mechanical cladding removal/flame cutting, Exothermic reaction, Diamond wire, Water jet, Laser, Mechanical milling, Controlled explosives, and Electrical discharge. After a detailed review of these 12 techniques, the decision was made by ANL that the most appropriate method for segmenting the EBWR reactor vessel would be to rift the vessel from the vessel cavity and use an abrasive water jet positioned on the main floor to perform the cutting of the reactor vessel

TORT application in reactor pressure vessel neutron flux calculations

International Nuclear Information System (INIS)

Belousov, S.I.; Ilieva, K.D.; Antonov, S.Y.

1994-01-01

The neutron flux values onto reactor pressure vessel for WWER-1000 and WWER-440 reactors, at the places important for metal embrittlement surveillance have been calculated by 3 dimensional code TORT and synthesis method. The comparison of the results received by both methods confirms their good consistency. (authors). 13 refs., 4 tabs

Heat insulation device for reactor pressure vessel in water

International Nuclear Information System (INIS)

Nakamura, Heiichiro; Tanaka, Yoshimi.

1993-01-01

Outer walls of a reactor pressure vessel are covered with water-tight walls made of metals. A heat insulation metal material is disposed between them. The water tight walls are joined by welding and flanges. A supply pipeline for filling gases and a discharge pipeline are in communication with the inside of the water tight walls. Further, a water detector is disposed in the midway of the gas discharge pipeline. With such a constitution, the following advantages can be attained. (1) Heat transfer from the reactor pressure vessel to water of a reactor container can be suppressed by filled gases and heat insulation metal material. (2) Since the pressure at the inside of the water tight walls can be equalized with the pressure of the inside of the reactor container, the thickness of the water-tight walls can be reduced. (3) Since intrusion of water to the inside of the walls due to rupture of the water tight walls is detected by the water detector, reactor scram can be conducted rapidly. (4) The sealing property of the flange joint portion is sufficient and detaching operation thereof is easy. (I.S.)

Effects of δ-hydride precipitation at a crack tip on crack propagation in delayed hydride cracking of Zircaloy-2

Energy Technology Data Exchange (ETDEWEB)

Kubo, T., E-mail: kubo@nfd.co.jp [Nippon Nuclear Fuel Development Co., Ltd., 2163 Narita-cho, Oarai-machi, Ibaraki 311-1313 (Japan); Kobayashi, Y. [M.O.X. Co., Ltd., 1828-520 Hirasu-cho, Mito, Ibaraki 311-0853 (Japan)

2013-08-15

Highlights: • Steady state crack velocity of delayed hydride cracking in Zircaloy-2 was analyzed. • A large stress peak is induced at an end of hydride by volume expansion of hydride. • Hydrogen diffuses to the stress peak, thereby accelerating steady hydride growth. • Crack velocity was estimated from the calculated hydrogen flux into the stress peak. • There was good agreement between calculation results and experimental data. -- Abstract: Delayed hydride cracking (DHC) of Zircaloy-2 is one possible mechanism for the failure of boiling water reactor fuel rods in ramp tests at high burnup. Analyses were made for hydrogen diffusion around a crack tip to estimate the crack velocity of DHC in zirconium alloys, placing importance on effects of precipitation of δ-hydride. The stress distribution around the crack tip is significantly altered by precipitation of hydride, which was strictly analyzed using a finite element computer code. Then, stress-driven hydrogen diffusion under the altered stress distribution was analyzed by a differential method. Overlapping of external stress and hydride precipitation at a crack tip induces two stress peaks; one at a crack tip and the other at the front end of the hydride precipitate. Since the latter is larger than the former, more hydrogen diffuses to the front end of the hydride precipitate, thereby accelerating hydride growth compared with that in the absence of the hydride. These results indicated that, after hydride was formed in front of the crack tip, it grew almost steadily accompanying the interaction of hydrogen diffusion, hydride growth and the stress alteration by hydride precipitation. Finally, crack velocity was estimated from the calculated hydrogen flux into the crack tip as a function of temperature, stress intensity factor and material strength. There was qualitatively good agreement between calculation results and experimental data.

Final report for the 2nd Ex-Vessel Neutron Dosimetry Installations and Evaluations for Yonggwang Unit 2 Reactor Pressure Vessel

Energy Technology Data Exchange (ETDEWEB)

Kim, Byoung Chul; Yoo, Choon Sung; Lee, Sam Lai; Gong, Un Sik; Choi, Kwon Jae; Chung, Kyoung Ki; Kim, Kwan Hyun; Chang, Jong Hwa; Ha, Jea Ju

2008-01-15

This report describes a neutron fluence assessment performed for the Yonggwang Unit 2 pressure vessel beltline region based on the guidance specified in Regulatory Guide 1.190. In this assessment, maximum fast neutron exposures expressed in terms of fast neutron fluence (E>1 MeV) and iron atom displacements (dpa) were established for the beltline region of the pressure vessel. During Cycle 16 of reactor operation, an Ex-Vessel Neutron Dosimetry Program was instituted at Yonggwang Unit 2 to provide continuous monitoring of the beltline region of the reactor vessel. The use of the Ex-Vessel Neutron Dosimetry Program coupled with available surveillance capsule measurements provides a plant specific data base that enables the evaluation of the vessel exposure and the uncertainty associated with that exposure over the service life of the unit. Ex-Vessel Neutron Dosimetry has been evaluated at the conclusion of Cycle 16.

Final report for the 1st ex-vessel neutron dosimetry installations and evaluations for Kori unit 2 reactor pressure vessel

Energy Technology Data Exchange (ETDEWEB)

Kim, Byoung Chul; Yoo, Choon Sung; Lee, Sam Lai; Chang, Kee Ok; Gong, Un Sik; Choi, Kwon Jae; Chang, Jong Hwa; Lim, Nam Jin; Hong, Joon Wha; Cheon, Byeong Jin

2006-11-15

This report describes a neutron fluence assessment performed for the Kori unit 2 pressure vessel belt line region based on the guidance specified in regulatory guide 1.190. In this assessment, maximum fast neutron exposures expressed in terms of fast neutron fluence (E>1 MeV) and iron atom displacements (dpa) were established for the belt line region of the pressure vessel. During cycle 20 of reactor operation, an ex-vessel neutron dosimetry program was instituted at Kori unit 2 to provide continuous monitoring of the belt line region of the reactor vessel. The use of the ex-vessel neutron dosimetry program coupled with available surveillance capsule measurements provides a plant specific data base that enables the evaluation of the vessel exposure and the uncertainty associated with that exposure over the service life of the unit. Ex-vessel neutron dosimetry has been evaluated at the conclusion of cycle 20.

Final Report of the 2nd Ex-Vessel Neutron Dosimetry Installation And Evaluations for Yonggwang Unit 1 Reactor Pressure Vessel

Energy Technology Data Exchange (ETDEWEB)

Kim, Byoung Chul; Yoo, Choon Sung; Lee, Sam Lai; Chang, Kee Ok; Gong, Un Sik; Choi, Kwon Jae; Chang, Jong Hwa; Li, Nam Jin; Hong, Joon Wha

2007-01-15

This report describes a neutron fluence assessment performed for the Yonggwang Unit 1 pressure vessel belt line region based on the guidance specified in Regulatory Guide 1.190. In this assessment, maximum fast neutron exposures expressed in terms of fast neutron fluence (E>1 MeV) and iron atom displacements (dpa) were established for the belt line region of the pressure vessel. During Cycle 16 of reactor operation, 2nd Ex-Vessel Neutron Dosimetry Program was instituted at Yonggwang Unit 1 to provide continuous monitoring of the belt line region of the reactor vessel. The use of the Ex-Vessel Neutron Dosimetry Program coupled with available surveillance capsule measurements provides a plant specific data base that enables the evaluation of the vessel exposure and the uncertainty associated with that exposure over the service life of the unit. Ex-Vessel Neutron Dosimetry has been evaluated at the conclusion of Cycle 16.

Final Report of the 2nd Ex-Vessel Neutron Dosimetry Installation And Evaluations for Yonggwang Unit 1 Reactor Pressure Vessel

Energy Technology Data Exchange (ETDEWEB)

Kim, Byoung Chul; Yoo, Choon Sung; Lee, Sam Lai; Gong, Un Sik; Choi, Kwon Jae; Chung, Kyoung Ki; Kim, Kwan Hyun; Chang, Jong Hwa; Ha, Jea Ju

2008-01-15

This report describes a neutron fluence assessment performed for the Kori Unit 2 pressure vessel belt line region based on the guidance specified in Regulatory Guide 1.190. In this assessment, maximum fast neutron exposures expressed in terms of fast neutron fluence (E>1 MeV) and iron atom displacements (dpa) were established for the belt line region of the pressure vessel. During Cycle 21 of reactor operation, an Ex-Vessel Neutron Dosimetry Program was instituted at Kori Unit 2 to provide continuous monitoring of the belt line region of the reactor vessel. The use of the Ex-Vessel Neutron Dosimetry Program coupled with available surveillance capsule measurements provides a plant specific data base that enables the evaluation of the vessel exposure and the uncertainty associated with that exposure over the service life of the unit. Ex-Vessel Neutron Dosimetry has been evaluated at the conclusion of Cycle 21.

Final report for the 1st ex-vessel neutron dosimetry installation and evaluations for Kori unit 4 reactor pressure vessel

Energy Technology Data Exchange (ETDEWEB)

Kim, Byoung Chul; Yoo, Choon Sung; Lee, Sam Lai; Chang, Kee Ok; Gong, Un Sik; Choi, Kwon Jae; Chang, Jong Hwa; Lim, Nam Jin; Hong, Joon Wha; Cheon, Byeong Jin

2006-11-15

This report describes a neutron fluence assessment performed for the Kori unit 4 pressure vessel belt line region based on the guidance specified in regulatory guide 1.190. In this assessment, maximum fast neutron exposures expressed in terms of fast neutron fluence (E>1 MeV) and iron atom displacements (dpa) were established for the belt line region of the pressure vessel. During cycle 16 of reactor operation, an ex-vessel neutron dosimetry program was instituted at Kori unit 4 to provide continuous monitoring of the belt line region of the reactor vessel. The use of the ex-vessel neutron dosimetry program coupled with available surveillance capsule measurements provides a plant specific data base that enables the evaluation of the vessel exposure and the uncertainty associated with that exposure over the service life of the unit. Ex-vessel neutron dosimetry has been evaluated at the conclusion of cycle 16.

Pressure vessel for a BWR type reactor

International Nuclear Information System (INIS)

Shimamoto, Yoshiharu.

1980-01-01

Purpose: To prevent the retention of low temperature water and also prevent the thermal fatigue of the pressure vessel by making large the curvature radius of a pressure vessel of a feed water sparger fitting portion and accelerating the mixing of low-temperature water at the feed water sparger base and in-pile hot water. Constitution: The curvature radius of the corner of the feed water sparger fitting portion in a pressure vessel is formed largely. In-pile circulating water infiltrates up to the base portion of the feed water sparger to carry outside low-temperature water at the base part, which is mixed with in-pile hot water. Accordingly, low temperature water does not stay at the base portion of the feed water sparger and generation of thermal fatigue in the pressure vessel can be prevented and the safety of the BWR type reactor can be improved. (Yoshino, Y.)

Study on the seismic response of reactor vessel of pool type LMFBR including fluid-structure interaction

International Nuclear Information System (INIS)

Tanimoto, K.; Ito, T.; Fujita, K.; Kurihara, C.; Sawada, Y.; Sakurai, A.

1988-01-01

The paper presents the seismic response of reactor vessel of pool type LMFBR with fluid-structure interaction. The reactor vessel has bottom support arrangement, the same core support system as Super-Phenix in France. Due to the bottom support arrangement, the level of core support is lower than that of the side support arrangement. So, in this reactor vessel, the displacement of the core top tends to increase because of the core's rocking. In this study, we investigated the vibration and seismic response characteristics of the reactor vessel. Therefore, the seismic experiments were carried out using one-eighth scale model and the seismic response including FSI and sloshing were investigated. From this study, the effect of liquid on the vibration characteristics and the seismic response characteristics of reactor vessel were clarified and sloshing characteristics were also clarified. It was confirmed that FEM analysis with FSI can reproduce the seismic behavior of the reactor vessel and is applicable to seismic design of the pool type LMFBR with bottom support arrangement. (author). 5 refs, 14 figs, 2 tabs

Development of transmutation technologies of radioactive waste by actinoid hydride

International Nuclear Information System (INIS)

Konashi, Kenji; Matsui, Hideki; Yamawaki, Michio

2001-01-01

Two waste treatment methods, geological disposal and transmutation, have been studied. The transmutation method changes long-lived radioactive nuclides to short-lived one or stabilizes them by nuclear transformation. The transmutation by actinoid hydride is exactly alike that transformation method from actinoid disposal waste to Pu fuel. For this object, OMEGA project is processing now. The transmutation is difficult by two causes such as large amount of long-lived radioactive nuclides and not enough development of control technologies of nuclear reaction except atomic reactor. The transmutation using actinoid hydride has merits that the amount of actinoid charged in the target increases and the effect of thermal neutrons on fuel decreases depending on homogeneous transmutation velocity in the target. Development of stable actinoid hydride under the conditions of reactor temperature and irradiation environment is important. The experimental results of U-ZrH 1.6 are shown in this paper. The irradiation experiment using Th hydride has been proceeding. (S.Y.)

Welding of the A1 reactor pressure vessel

International Nuclear Information System (INIS)

Becka, J.

1975-01-01

As concerns welding, the A-1 reactor pressure vessel represents a geometrically complex unit containing 1492 welded joints. The length of welded sections varies between 10 and 620 mm. At an operating temperature of 120 degC and a pressure of 650 N/cm 2 the welded joints in the reactor core are exposed to an integral dose of 3x10 18 n/cm 2 . The chemical composition is shown for pressure vessel steel as specified by CSN 413090.9 modified by Ni, Ti and Al additions, and for the welding electrodes used. The requirements are also shown for the mechanical properties of the base and the weld metals. The technique and conditions of welding are described. No defects were found in ultrasonic testing of welded joints. (J.B.)

Load bearing capacities and elastic-plastic behavior of reactor vessel internals

International Nuclear Information System (INIS)

Watanabe, Keita; Nagase, Ryuichi

2017-01-01

Radial Support Keys (RSKs) are installed at the bottom of Reactor Vessel Internal (RVI) of Pressurized Water Reactor (PWR) and fit into Core Support Lugs of Reactor Pressure Vessel (RPV). This structure provides reactor core horizontal support and transmits the loads between RVI and RPV. RSK is one of the critical parts of RVI from the view point of earthquake-proof safety. In order to assure the structural integrity of Nuclear Reactor in case of massive earthquake, load bearing capacities of RSK are confirmed by static loading tests with reduced-scale mockups. In addition, collapse loads of actual components calculated by Limit Analyses are conservative enough compared to the load bearing capacities confirmed by the test. Thus, the methodology to calculate collapse load by Limit Analysis is applicable to evaluation of structural integrity for RSK. (author)

Performance demonstration experience for reactor pressure vessel shell ultrasonic testing

International Nuclear Information System (INIS)

Zado, V.

1998-01-01

The most ultrasonic testing techniques used by many vendors for pressurized water reactor (PWR) examinations were based on American Society of Mechanical Engineers 'Boiler and Pressurized Vessel Code' (ASME B and PV Code) Sections XI and V. The Addenda of ASME B and PV Code Section XI, Edition 1989 introduced Appendix VIII - 'Performance Demonstration for Ultrasonic Examination Systems'. In an effort to increase confidence in performance of ultrasonic testing of the operating nuclear power plants in United States, the ultrasonic testing performance demonstration examination of reactor vessel welds is performed in accordance with Performance Demonstration Initiative (PDI) program which is based on ASME Code Section XI, Appendix VIII requirements. This article provides information regarding extensive qualification preparation works performed prior EPRI guided performance demonstration exam of reactor vessel shell welds accomplished in January 1997 for the scope of Appendix VIII, Supplements IV and VI. Additionally, an overview of the procedures based on requirements of ASME Code Section XI and V in comparison to procedure prepared for Appendix VIII examination is given and discussed. The samples of ultrasonic signals obtained from artificial flaws implanted in vessel material are presented and results of ultrasonic testing are compared to actual flaw sizes. (author)

Consequence evaluation of hypothetical reactor pressure vessel support failure

International Nuclear Information System (INIS)

Lu, S.C.; Holman, G.S.; Lambert, H.E.

1991-01-01

This paper describes a consequence evaluation to address safety concerns raised by the radiation embrittlement of the reactor pressure vessel (RPV) supports for the Trojan nuclear power plant. The study comprises a structural evaluation and an effects evaluation and assumes that all four reactor vessel supports have completely lost the load carrying capability. The structural evaluation concludes that the Trojan reactor coolant loop (RCL) piping is capable of transferring loads to the steam generator (SG) supports and the reactor coolant pump (RCP) supports and that the SG supports and the RCP supports have sufficient design margins to accommodate additional loads transferred to them through the RCL piping. The effects evaluation, employing a systems analysis approach, investigates initiating events and the reliability of the engineered safeguard systems as the RPV is subject to movements caused by the RPV support failure. The evaluation identifies a number of areas for further investigation and concludes that a hypothetical failure of the Trojan RPV supports due to radiation embrittlement will not result in consequences of significant safety concerns. (author)

Finite element analysis of thermal stresses of the reactor vessel in a severe light water reactor accident

International Nuclear Information System (INIS)

Borovkov, A.I.; Semenov, A.S.; Granovsky, V.S.; Kovtunova, S.V.

1995-01-01

The thermal stress and damage analysis of the light water reactor (LWR) vessel is considered in a severe accident conditions. The high temperature corium accumulates on the vessel bottom and necessary condition of its holding is intensive cooling of vessel. External flooding with outside cooling of the LWR vessel is one of the accident management strategies being proposed to ensure the integrity of the vessel after a severe accident. (author). 8 refs., 5 figs

Finite element analysis of thermal stresses of the reactor vessel in a severe light water reactor accident

Energy Technology Data Exchange (ETDEWEB)

Borovkov, A.I.; Semenov, A.S. [St. Petersburg State Technical Univ. (Russian Federation); Granovsky, V.S.; Kovtunova, S.V. [Research Inst. of Technology, Sosnovy Bor (Russian Federation)

1995-12-31

The thermal stress and damage analysis of the light water reactor (LWR) vessel is considered in a severe accident conditions. The high temperature corium accumulates on the vessel bottom and necessary condition of its holding is intensive cooling of vessel. External flooding with outside cooling of the LWR vessel is one of the accident management strategies being proposed to ensure the integrity of the vessel after a severe accident. (author). 8 refs., 5 figs.

Requirements for thermal insulation on prestressed concrete reactor vessels

International Nuclear Information System (INIS)

Neylan, A.J.; Wistrom, J.D.

1979-01-01

During the past decade, extensive design, construction, and operating experience on concrete pressure vessels for gas-cooled reactor applications has accumulated. Excellent experience has been obtained to date on the structural components (concrete, prestressing systems, liners, penetrations, and closures) and the thermal insulation. Three fundamentally different types of insulation systems have been employed to ensure the satisfactory performance of this component, which is critical to the overall success of the prestressed concrete reactor vessel (PCRV). Although general design criteria have been published, the requirements for design, materials, and construction are not rigorously addressed in any national or international code. With the more onerous design conditions being imposed by advanced reactor systems, much greater attention has been directed to advance the state of the art of insulation systems for PCRVs. This paper addresses some of the more recent developments in this field being performed by General Atomic Company and others. (author)

Effect of radiation damage on operating safety of steel pressure vessels of nuclear reactors

International Nuclear Information System (INIS)

Vacek, M.; Havel, S.; Stoces, B.; Brumovsky, M.

1980-01-01

The effects are assessed of the environment upon mechanical properties of steel used generally for pressure vessels of light water nuclear reactors. Changes caused by radiation affect the reliability of vessels. Deterioration of steel properties is mainly due to neutron radiation. The article deals with factors bearing upon damage and with methods allowing to evaluate the reliability of vessels and predict their service life. Operating reliability of vessels is very unfavourably affected by planned and accidental reactor transients. (author)

Superheated steam annealing of pressurized water reactor vessel

International Nuclear Information System (INIS)

Porowski, J.S.

1993-01-01

Thermal annealing of an embrittled Reactor Pressure Shell is the only recognized means for recovering material properties lost due to long-term exposure of the reactor walls to radiation. Reduced toughness of the material during operation is a major concern in evaluations of structural integrity of older reactors. Extensive studies performed within programs related to life extension of nuclear plants have confirmed that the thermal treatment of 850 deg. F for 168 hours on irradiated material essentially recovers material properties lost due to neutron exposure. Dry and wet annealing methods have been considered. Wet annealing involves operating the reactor at near design temperatures and pressures. Since the temperature of wet annealing must be limited to vessel design temperature of 650 deg. F, only partial recovery of the lost properties is achieved. Thus dry annealing was selected as an alternative for future development and industrial implementation to extend the safe life of reactors. Dry thermal annealing consists of heating portions of the reactor vessel at a specific temperature for a given period of time using a high temperature heat source. The use of spent fuel assemblies, induction heating and resistance heating elements as well as the circulation of heated fluid were investigated as potential candidate methods. To date the use of resistance heating elements which are lowered into a dry empty reactor was considered to be the preferred method. In-depth research in the United States and practical applications of such a method in Russia have confirmed feasibility of the method. The method of using circulating superheated steam to anneal the vessel at 850 deg. F without complete removal of the reactor internals is described herein. After removing the reactor head and fuel, the core barrel along with the upper and lower core in PWRs is lifted to open an annular space between the reactor shell flange and the core barrel flange. The thermal shield can remain

Underwater laser beam welding technology for reactor vessel nozzles of PWRs

International Nuclear Information System (INIS)

Yoda, Masaki; Tamura, Masataka; Tamura, Masataka

2010-01-01

Toshiba has developed an underwater laser beam welding technology for the maintenance of reactor vessel nozzles of pressurized water reactors (PWRs), which eliminates the need for the drainage of water from the reactor vessel. The new welding system makes it possible to both reduce the work period and minimize the radiation exposure of workers compared with conventional technologies for welding in ambient air. We have confirmed the effectiveness of this technology through experiments in which stress corrosion cracking (SCC) was mitigated on the inner surfaces of nozzles. We are promoting its practical application in Japan and overseas in cooperation with Westinghouse Electric Company, a group company of Toshiba. (author)

1-D Two-phase Flow Investigation for External Reactor Vessel Cooling

International Nuclear Information System (INIS)

Kim, Jae Cheol

2007-02-01

During a severe accident, when a molten corium is relocated in a reactor vessel lower head, the RCF(Reactor Cavity Flooding) system for ERVC (External Reactor Vessel Cooling) is actuated and coolants are supplied into a reactor cavity to remove a decay heat from the molten corium. This severe accident mitigation strategy for maintaining a integrity of reactor vessel was adopted in the nuclear power plants of APR1400, AP600, and AP1000. Under the ERVC condition, the upward two-phase flow is driven by the amount of the decay heat from the molten corium. To achieve the ERVC strategy, the two-phase natural circulation in the annular gap between the external reactor vessel and the insulation should be formed sufficiently by designing the coolant inlet/outlet area and gap size adequately on the insulation device. Also the natural circulation flow restriction has to be minimized. In this reason, it is needed to review the fundamental structure of insulation. In the existing power plants, the insulation design is aimed at minimizing heat losses under a normal operation. Under the ERVC condition, however, the ability to form the two-phase natural circulation is uncertain. Namely, some important factors, such as the coolant inlet/outlet areas, flow restriction, and steam vent etc. in the flow channel, should be considered for ERVC design. T-HEMES 1D study is launched to estimate the natural circulation flow under the ERVC condition of APR1400. The experimental facility is one-dimensional and scaled down as the half height and 1/238 channel area of the APR1400 reactor vessel. The air injection method was used to simulate the boiling at the external reactor vessel and generate the natural circulation two-phase flow. From the experimental results, the natural circulation flow rate highly depended on inlet/outlet areas and the circulation flow rate increased as the outlet height as well as the supplied water head increased. On the other hand, the simple analysis using the drift

Large-scale testing of in-vessel debris cooling through external flooding of the reactor pressure vessel in the CYBL facility

International Nuclear Information System (INIS)

Chu, T.Y.; Bentz, J.H.; Bergeron, K.D.; Slezak, S.E.; Simpson, R.B.

1994-01-01

The possibility of achieving in-vessel core retention by flooding the reactor cavity, or the ''flooded cavity'', is an accident management concept currently under consideration for advanced light water reactors (ALWR), as well as for existing light water reactors (LWR). The CYBL (CYlindrical BoiLing) facility is a facility specifically designed to perform large-scale confirmatory testing of the flooded cavity concept. CYBL has a tank-within-a-tank design; the inner 3.7 m diameter tank simulates the reactor vessel, and the outer tank simulates the reactor cavity. The energy deposition on the bottom head is simulated with an array of radiant heaters. The array can deliver a tailored heat flux distribution corresponding to that resulting from core melt convection. The present paper provides a detailed description of the capabilities of the facility, as well as results of recent experiments with heat flux in the range of interest to those required for in-vessel retention in typical ALWRs. The paper concludes with a discussion of other experiments for the flooded cavity applications

Large-Scale testing of in-vessel debris cooling through external flooding of the reactor pressure vessel in the CYBL facility

International Nuclear Information System (INIS)

Chu, T.Y.; Bentz, J.H.; Bergeron, K.D.; Slezak, S.E.; Simpson, R.B.

1994-01-01

The possibility of achieving in-vessel core retention by flooding the reactor cavity, or the open-quotes flooded cavityclose quotes, is an accident management concept currently under consideration for advanced light water reactors (ALWR), as well as for existing light water reactors (LWR). The CYBL (CYlindrical BoiLing) facility is a facility specifically designed to perform large-scale confirmatory testing of the flooded cavity concept. CYBL has a tank-within-a-tank design; the inner 3.7 m diameter tank simulates the reactor vessel, and the outer tank simulates the reactor cavity. The energy deposition on the bottom head is simulated with an array of radiant heaters. The array can deliver a tailored heat flux distribution corresponding to that resulting from core melt convection. The present paper provides a detailed description of the capabilities of the facility, as well as results of recent experiments with heat flux in the range of interest to those required for in-vessel retention in typical ALWRs. The paper concludes with a discussion of other experiments for the flooded cavity applications

Vessel supporting structure for liquid metal cooled nuclear reactors

International Nuclear Information System (INIS)

Mahe, Armel; Jullien, Georges

1974-01-01

The supporting structure described is for a liquid metal cooled nuclear reactor, the vessel being of the type suspended to the end slab of the reactor. It includes a ring connected at one of its two ends to a single shell and at the other end to two shells. One of these three shells connected to the lower end of the ring forms the upper part of the vessel to be supported. The two other shells are embedded in two sperate parts of the slab. The ring and shell assembly is housed in an annular space provided in the end slab and separating it into two parts, namely a central part and a peripheral part [fr

Baffle-former arrangement for nuclear reactor vessel internals

International Nuclear Information System (INIS)

Machado, O.J.; Berringer, R.T.

1978-01-01

A baffle-former arrangement for the reactor vessel internals of a nuclear reactor is described. The arrangement includes positioning of formers at the same elevations as the fuel assembly grids, and positioning flow holes in the baffle plates directly beneath selected former grid elevations. The arrangement reduces detrimental cross flows, maintains proper core barrel and baffle temperatures, and alleviates the potential of overpressurization within the baffle-former assembly under assumed major accident conditions

Case study for one-piece removal method of reactor vessel of nuclear ship 'Mutsu'

International Nuclear Information System (INIS)

Nagane, Satoru; Kitahara, Katsumi; Yoshikawa, Seiji; Miyasaka, Yasuhiko; Fukumura, Nobuo; Nisizawa, Ichiou

2010-01-01

A reactor installed at the center part of the nuclear ship 'Mutsu' has been stored safely and exhibited in a reactor room building since 1996. The reactor vessel and its internals are key components because of main radioactive wastes for the reasonable decommissioning plan in the future. This report describes the one-piece removal method as the one package of the reactor vessel with its internals intact with a shipping container or additional shields. The reactor vessel package (Max.100ton) will be classified acceptable for burial at the low level radioactive waste (LLW), which will be buried at a LLW pit facility under waste disposal regulations. And also, the package will be classified as an IP-2-equivalent package according to the requirement for Shipments and Packagings. (author)

Internal hydriding in irradiated defected Zircaloy fuel rods: A review (LWBR Development Program)

International Nuclear Information System (INIS)

Clayton, J.C.

1987-10-01

Although not a problem in recent commercial power reactors, including the Shippingport Light Water Breeder Reactor, internal hydriding of Zircaloy cladding was a persistent cause of gross cladding failures during the 1960s. It occurred in the fuel rods of water-cooled nuclear power reactors that had a small cladding defect. This report summarizes the experimental findings, causes, mechanisms, and methods of minimizing internal hydriding in defected Zircaloy-clad fuel rods. Irradiation test data on the different types of defected fuel rods, intentionally fabricated defected and in-pile operationally defected rods, are compared. Significant factors affecting internal hydriding in defected Zircaloy-clad fuel rods (defect hole size, internal and external sources of hydrogen, Zircaloy cladding surface properties, nickel alloy contamination of Zircaloy, the effect of heat flux and fluence) are discussed. Pertinent in-pile and out-of-pile test results from Bettis and other laboratories are used as a data base in constructing a qualitative model which explains hydrogen generation and distribution in Zircaloy cladding of defected water-cooled reactor fuel rods. Techniques for minimizing internal hydride failures in Zircaloy-clad fuel rods are evaluated

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

International Nuclear Information System (INIS)

1998-10-01

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

Lessons Learned From Developing Reactor Pressure Vessel Steel Embrittlement Database

Energy Technology Data Exchange (ETDEWEB)

Wang, Jy-An John [ORNL

2010-08-01

Materials behaviors caused by neutron irradiation under fission and/or fusion environments can be little understood without practical examination. Easily accessible material information system with large material database using effective computers is necessary for design of nuclear materials and analyses or simulations of the phenomena. The developed Embrittlement Data Base (EDB) at ORNL is this comprehensive collection of data. EDB database contains power reactor pressure vessel surveillance data, the material test reactor data, foreign reactor data (through bilateral agreements authorized by NRC), and the fracture toughness data. The lessons learned from building EDB program and the associated database management activity regarding Material Database Design Methodology, Architecture and the Embedded QA Protocol are described in this report. The development of IAEA International Database on Reactor Pressure Vessel Materials (IDRPVM) and the comparison of EDB database and IAEA IDRPVM database are provided in the report. The recommended database QA protocol and database infrastructure are also stated in the report.

Analysis of stress in reactor core vessel under effect of pressure lose shock wave

International Nuclear Information System (INIS)

Li Yong; Liu Baoting

2001-01-01

High Temperature gas cooled Reactor (HTR-10) is a modular High Temperature gas cooled Reactor of the new generation. In order to analyze the safety characteristics of its core vessel in case of large rupture accident, the transient performance of its core vessel under the effect of pressure lose shock wave is studied, and the transient pressure difference between the two sides of the core vessel and the transient stresses in the core vessel is presented in this paper, these results can be used in the safety analysis and safety design of the core vessel of HTR-10. (author)

Nonlinear analysis of end slabs in prestressed concrete reactor vessels

International Nuclear Information System (INIS)

Abdulrahman, H.O.

1978-01-01

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

Materials for high temperature reactor vessels

International Nuclear Information System (INIS)

Buenaventura Pouyfaucon, A.

2004-01-01

Within the 5th Euraton Framework Programme, a big effort is being made to promote and consolidate the development of the High Temperature Reactor (HTR). Empresarios Agrupados is participating in this project and among others, also forms part of the HTR-M project Materials for HTRs. This paper summarises the work carried out by Empresarios Agrupados regarding the material selection of the HTR Reactor Pressure Vessel (RPV). The possible candidate materials and the most promising ones are discussed. Design aspects such as the RPV sensitive zones and material damage mechanisms are considered. Finally, the applicability of the existing design Codes and Standards for the design of the HTR RPV is also discussed. (Author)

Accelerated irradiation test of Gundremmingen reactor vessel trepan material

Energy Technology Data Exchange (ETDEWEB)

Hawthorne, J.R. [Materials Engineering Associates, Inc., Lanham, MD (United States)

1992-08-01

Initial mechanical properties tests of beltline trepanned from the decommissioned KRB-A pressure vessel and archive material irradiated in the UBR test reactor revealed a major anomaly in relative radiation embrittlement sensitivity. Poor correspondence of material behavior in test vs. power reactor environments was observed for the weak test orientation (ASTL C-L) whereas correspondence was good for the strong orientation (ASTM C-L). To resolve the anomaly directly, Charpy-V specimens from a low (essentially-nil) fluence region of the vessel were irradiated together with archive material at 279{degrees}C in the UBR test reactor. Properties tests before UBR irradiation revealed a significant difference in 41-J transition temperature and upper shelf energy level between the materials. However, the materials exhibited essentially the same radiation embrittlement sensitivity (both orientations), proving that the anomaly is not due to a basic difference in material irradiation resistances. Possible causes of the original anomaly and the significance to NRC Regulatory Guide 1.99 are discussed.

Accelerated irradiation test of gundremmingen reactor vessel trepan material

International Nuclear Information System (INIS)

Hawthorne, J.R.

1992-08-01

Initial mechanical properties tests of beltline trepanned from the decommissioned KRB-A pressure vessel and archive material irradiated in the UBR test reactor revealed a major anomaly in relative radiation embrittlement sensitivity. Poor correspondence of material behavior in test vs. power reactor environments was observed for the weak test orientation (ASTL C-L) whereas correspondence was good for the strong orientation (ASTM C-L). To resolve the anomaly directly, Charpy-V specimens from a low (essentially-nil) fluence region of the vessel were irradiated together with archive material at 279 degrees C in the UBR test reactor. Properties tests before UBR irradiation revealed a significant difference in 41-J transition temperature and upper shelf energy level between the materials. However, the materials exhibited essentially the same radiation embrittlement sensitivity (both orientations), proving that the anomaly is not due to a basic difference in material irradiation resistances. Possible causes of the original anomaly and the significance to NRC Regulatory Guide 1.99 are discussed

Stresses in reactor pressure vessel nozzles -- Calculations and experiments

International Nuclear Information System (INIS)

Brumovsky, M.; Polachova, H.

1995-01-01

Reactor pressure vessel nozzles are characterized by a high stress concentration which is critical in their low-cycle fatigue assessment. Program of experimental verification of stress/strain field distribution during elastic-plastic loading of a reactor pressure vessel WWER-1000 primary nozzle model in scale 1:3 is presented. While primary nozzle has an ID equal to 850 mm, the model nozzle has ID equal to 280 mm, and was made from 15Kh2NMFA type of steel. Calculation using analytical methods was performed. Comparison of results using different analytical methods -- Neuber's, Hardrath-Ohman's as well as equivalent energy ones, used in different reactor Codes -- is shown. Experimental verification was carried out on model nozzles loaded statically as well as by repeated loading, both in elastic-plastic region. Strain fields were measured using high-strain gauges, which were located in different distances from center of nozzle radius, thus different stress concentration values were reached. Comparison of calculated and experimental data are shown and compared

Experiments and analysis of thermal stresses around the nozzle of the reactor vessel

International Nuclear Information System (INIS)

Song, D.H.; Oh, J.H.; Song, H.K.; Park, D.S.; Shon, K.H.

1981-01-01

This report describes the results of analysis and experiments on the thermal stress around the reactor vessel nozzle performed to establish a capability of thermal stress analysis of pressure vessel subjected to thermal loadings. Firstly, heat conduction analysis during reactor design transients and analysis on the experimental model were performed using computer code FETEM-1 for the purpose of verification of FETEM-1 which was developed in 1979 and will be used to obtain the temperature distribution in a solid body under the steady-state and the transient conditions. The results of the analysis was compared to the results in the Stress Report of Kori-1 reactor vessel and those from experiments on the model, respectively

Swedish Work on Brittle-Fracture Problems in Nuclear Reactor Pressure Vessels

Energy Technology Data Exchange (ETDEWEB)

Grounes, M

1966-03-15

After a short review of the part of the Swedish nuclear energy program that is of interest in this context the Swedish reactor pressure vessels and the reasoning behind the choice of materials are surveyed. Problems and desirable aims for future reactors are discussed. Much work is now being done on new types of pressure vessel steels with high strength, low transition temperature and good corrosion resistance. These steels are of the martensitic austenitic type Bofors 2RMO (13 % Cr, 6 % Ni, 1. 5 % Mo) and of the ferritic martensitic austenitic type Avesta 248 SV (16 % Cr, 5 % Ni, 1 % Mo). An applied philosophy for estimating the brittle-fracture tendency of pressure vessels is described. As a criterion of this tendency we use the crack-propagation transition temperature, e. g. as measured by the Robertson isothermal crack-arrest test. An estimate of this transition temperature at the end of the reactor' s lifetime must take increases due to fabrication, welding, geometry, ageing and irradiation into account. The transition temperature vs. stress curve moves towards higher temperatures during the reactor' s lifetime. As long as this curve does not cross the reactor vessel stress vs. temperature curve the vessel is considered safe. The magnitude of the different factors influencing the final transition temperature are discussed and data for the Marviken reactor's pressure vessel are presented. At the end of the reactor's lifetime the estimated transition temperature is 115 deg C, which is below the maximum permissible value. A program for the study of strain ageing has been initiated owing to the uncertainty as to the extent of strain ageing at low strains. A study of a simple crack-arrest test, developed in Sweden, is in progress. An extensive irradiation-effects program on several steels is in progress. Results from tests on the Swedish carbon-manganese steels 2103/R3, SIS 142103 and SIS 142102, the low-alloy steels Degerfors DE-631A, Bofors NO 345 and Fortiweld

Swedish Work on Brittle-Fracture Problems in Nuclear Reactor Pressure Vessels

International Nuclear Information System (INIS)

Grounes, M.

1966-03-01

After a short review of the part of the Swedish nuclear energy program that is of interest in this context the Swedish reactor pressure vessels and the reasoning behind the choice of materials are surveyed. Problems and desirable aims for future reactors are discussed. Much work is now being done on new types of pressure vessel steels with high strength, low transition temperature and good corrosion resistance. These steels are of the martensitic austenitic type Bofors 2RMO (13 % Cr, 6 % Ni, 1. 5 % Mo) and of the ferritic martensitic austenitic type Avesta 248 SV (16 % Cr, 5 % Ni, 1 % Mo). An applied philosophy for estimating the brittle-fracture tendency of pressure vessels is described. As a criterion of this tendency we use the crack-propagation transition temperature, e. g. as measured by the Robertson isothermal crack-arrest test. An estimate of this transition temperature at the end of the reactor' s lifetime must take increases due to fabrication, welding, geometry, ageing and irradiation into account. The transition temperature vs. stress curve moves towards higher temperatures during the reactor' s lifetime. As long as this curve does not cross the reactor vessel stress vs. temperature curve the vessel is considered safe. The magnitude of the different factors influencing the final transition temperature are discussed and data for the Marviken reactor's pressure vessel are presented. At the end of the reactor's lifetime the estimated transition temperature is 115 deg C, which is below the maximum permissible value. A program for the study of strain ageing has been initiated owing to the uncertainty as to the extent of strain ageing at low strains. A study of a simple crack-arrest test, developed in Sweden, is in progress. An extensive irradiation-effects program on several steels is in progress. Results from tests on the Swedish carbon-manganese steels 2103/R3, SIS 142103 and SIS 142102, the low-alloy steels Degerfors DE-631A, Bofors NO 345 and Fortiweld

Surveillance of irradiation embrittlement of nuclear reactor pressure vessels

International Nuclear Information System (INIS)

Najzer, M.

1982-01-01

Surveillance of irradiation embrittlement of nuclear reactor pressure vessels is briefly discussed. The experimental techniques and computer programs available for this work at the J. Stefan Institute are described. (author)

Review of consequences of uranium hydride formation in N-Reactor fuel elements stored in the K-Basins

Energy Technology Data Exchange (ETDEWEB)

Weber, J.W.

1994-09-28

The 105-K Basins on the Hanford site are used to store uranium fuel elements and assemblies irradiated in and discharged from N Reactor. The storage cylinders in KW Basin are known to have some broken N reactor fuel elements in which the exposed uranium is slowly reacting chemically with water in the cylinder. The products of these reactions are uranium oxide, hydrogen, and potentially some uranium hydride. The purpose of this report is to document the results f the latest review of potential, but highly unlikely accidents postulated to occur as closed cylinders containing N reactor fuel assemblies are opened under water in the KW basin and as a fuel assembly is raised from the basin in a shipping cask for transportation to the 327 Building for examination as part of the SNF Characterization Program. The postulated accidents reviews in this report are considered to bound all potential releases of radioactivity and hydrogen. These postulated accidents are: (1) opening and refill of a cylinder containing significant amounts of hydrogen and uranium hydride; and (2) draining of the single element can be used to keep the fuel element submerged in water after the cask containing the can and element is lifted from the KW Basin. Analysis shows the release of radioactivity to the site boundary is significantly less than that allowed by the K Basin Safety Evaluation. Analysis further shows there would be no damage to the K Basin structure nor would there be injury to personnel for credible events.

The application of probabilistic fracture analysis to residual life evaluation of embrittled reactor vessels

International Nuclear Information System (INIS)

Dickson, T.L.; Simonen, F.A.

1992-01-01

Probabilistic fracture mechanics analysis is a major element of the comprehensive probabilistic methodology on which current NRC regulatory requirements for pressurized water reactor vessel integrity evaluation are based. Computer codes such as OCA-P and VISA-11 perform probabilistic fracture analyses to estimate the increase in vessel failure probability that occurs as the vessel material accumulates radiation damage over the operating life of the vessel. The results of such analyses, when compared with limits of acceptable failure probabilities, provide an estimation of the residual life of a vessel. Such codes can be applied to evaluate the potential benefits of plant-specific mitigating actions designed to reduce the probability of failure of a reactor vessel

The application of probabilistic fracture analysis to residual life evaluation of embrittled reactor vessels

International Nuclear Information System (INIS)

Dickson, T.L.; Simonen, F.A.

1992-01-01

Probabilistic fracture mechanics analysis is a major element of comprehensive probabilistic methodology on which current NRC regulatory requirements for pressurized water reactor vessel integrity evaluation are based. Computer codes such as OCA-P and VISA-II perform probabilistic fracture analyses to estimate the increase in vessel failure probability that occurs as the vessel material accumulates radiation damage over the operating life of the vessel. The results of such analyses, when compared with limits of acceptable failure probabilities, provide an estimation of the residual life of a vessel. Such codes can be applied to evaluate the potential benefits of plant-specific mitigating actions designed to reduce the probability of failure of a reactor vessel. 10 refs

Strain measurement in and analysis for hydraulic test of CPR1000 reactor pressure vessel

International Nuclear Information System (INIS)

Zhou Dan; Zhuang Dongzhen

2013-01-01

The strain measurement in hydraulic test of CPR1000 reactor pressure vessel performed in Dongfang Heavy Machinery Co., Ltd. is introduced. The detail test scheme and method was introduced and the measurement results of strain and stress was given. Meanwhile the finite element analysis was performed for the pressure vessel, which was generally matched with the measurement results. The reliability of strain measurement was verified and the high strength margin of vessel was shown, which would give a good reference value for the follow-up hydraulic tests and strength analysis of reactor pressure vessel. (authors)

Pressure Tube and Pressure Vessel Reactors; certain comparisons

Energy Technology Data Exchange (ETDEWEB)

Margen, P H; Ahlstroem, P E; Pershagen, B

1961-04-15

In a comparison between pressure tube and pressure vessel type reactors for pressurized D{sub 2}O coolant and natural uranium, one can say that reactors of these two types having the same net electrical output, overall thermal efficiency, reflected core volume and fuel lattice have roughly the same capital cost. In these circumstances, the fuel burn-up obtainable has a significant influence on the relative economics. Comparisons of burn-up values made on this basis are presented in this report and the influence on the results of certain design assumptions are discussed. One of the comparisons included is based on the dimensions and ratings proposed for CANDU. Moderator temperature coefficients are compared and differences in kinetic behaviour which generally result in different design philosophies for the two types are mentioned, A comparison of different methods of obtaining flux flattening is presented. The influence of slight enrichment and other coolants, (boiling D{sub 2}O and gases) on the comparison between pressure tube and pressure vessel designs is discussed and illustrated with comparative designs for 400 MW electrical output. This paper was presented at the EAES Enlarged Symposium on Heterogeneous Heavy Water Power Reactors, Mallorca, October 10 - 14, 1960.

Pressure Tube and Pressure Vessel Reactors; certain comparisons

International Nuclear Information System (INIS)

Margen, P.H.; Ahlstroem, P.E.; Pershagen, B.

1961-04-01

In a comparison between pressure tube and pressure vessel type reactors for pressurized D 2 O coolant and natural uranium, one can say that reactors of these two types having the same net electrical output, overall thermal efficiency, reflected core volume and fuel lattice have roughly the same capital cost. In these circumstances, the fuel burn-up obtainable has a significant influence on the relative economics. Comparisons of burn-up values made on this basis are presented in this report and the influence on the results of certain design assumptions are discussed. One of the comparisons included is based on the dimensions and ratings proposed for CANDU. Moderator temperature coefficients are compared and differences in kinetic behaviour which generally result in different design philosophies for the two types are mentioned, A comparison of different methods of obtaining flux flattening is presented. The influence of slight enrichment and other coolants, (boiling D 2 O and gases) on the comparison between pressure tube and pressure vessel designs is discussed and illustrated with comparative designs for 400 MW electrical output. This paper was presented at the EAES Enlarged Symposium on Heterogeneous Heavy Water Power Reactors, Mallorca, October 10 - 14, 1960

Comparison of In-Vessel Shielding Design Concepts between Sodium-cooled Fast Burner Reactor and the Sodium-cooled Fast Breeder Reactor

International Nuclear Information System (INIS)

Yun, Sunghwan; Kim, Sang Ji

2015-01-01

In this study, quantities of in-vessel shields were derived and compared each other based on the replaceable shield assembly concept for both of the breeder and burner SFRs. Korean Prototype Gen-IV Sodium-cooled Fast Reactor (PGSFR) like SFR was used as the reference reactor and calculation method reported in the reference was used for shielding analysis. In this paper, characteristics of in-vessel shielding design were studied for the burner SFR and breeder SFR based on the replaceable shield assembly concept. An in-vessel shield to prevent secondary sodium activation (SSA) in the intermediate heat exchangers (IHXs) is one of the most important structures for the pool type Sodium-cooled Fast Reactor (SFR). In our previous work, two in-vessel shielding design concepts were compared each other for the burner SFR. However, a number of SFRs have been designed and operated with the breeder concept, in which axial and radial blankets were loaded for fuel breeding, during the past several decades. Since axial and radial blanket plays a role of neutron shield, comparison of required in-vessel shield amount between the breeder and burner SFRs may be an interesting work for SFR designer. Due to the blanket, the breeder SFR showed better performance in axial neutron shielding. Hence, 10.1 m diameter reactor vessel satisfied the design limit of SSA at the IHXs. In case of the burner SFR, due to more significant axial fast neutron leakage, 10.6 m diameter reactor vessel was required to satisfy the design limit of SSA at the IHXs. Although more efficient axial shied such as a mixture of ZrH 2 and B 4 C can improve shielding performance of the burner SFR, additional fabrication difficulty may mitigate the advantage of improved shielding performance. Therefore, it can be concluded that the breeder SFR has better characteristic in invessel shielding design to prevent SSA at the IHXs than the burner SFR in the pool-type reactor

Design system for in-vessel mainipulator of fusion reactor 'DESIM'

International Nuclear Information System (INIS)

Adachi, Junihci; Kobayashi, Takeshi; Ise, Hideo; Sato, Keisuke; Matsuda, Hirotsugu

1989-01-01

A computer aided design system 'DESIM' for the in-vessel manipulators of nuclear fusion reactors has been developed to design the manipulators efficiently. The DESIM consists of the following subsystems: (1) the design system for arm mechanisms to realize optimum manipulation performance in the specified workspace; (2) the robot simulator to study manipulator movement, postures and interference problems; (3) the CAD system which is used to define the structure object data for robots, and the interface system for the data conversion from the CAD system to the robot simulator. The DESIM has been used to design the in-vessel manipulator for the Fusion Experimental Reactor (FER) to confirm the effectiveness. (author)

Debris interactions in reactor vessel lower plena during a severe accident. II. Integral analysis

International Nuclear Information System (INIS)

Suh, K.Y.; Henry, R.E.

1996-01-01

For pt.I see ibid., p.147-63, 1996. The integral physico-numerical model for the reactor vessel lower head response has been exercised for the TMI-2 accident and possible severe accident scenarios in PWR and BWR designs. The proposed inherent cooling mechanism of the reactor material creep and subsequent water ingression implemented in this predictive model provides a consistent representation of how the debris was finally cooled in the TMI-2 accident and how the reactor lower head integrity was maintained during the course of the incident. It should be recalled that in order for this strain to occur, the vessel lower head had to achieve temperatures in excess of 1000 C. This is certainly in agreement with the temperatures determined by metallographic examinations during the TMI-2 vessel inspection program. The integral model was also applied to typical PWR and BWR lower plena with and without structures under pressurized conditions spanning the first relocation of core material to the reactor vessel failure due to creep without recovery actions. The design application results are presented with particular attention being focused on water ingression into the debris bed through the gap formed between the debris and the vessel wall. As an illustration of the accident management application, the lower plenum with structures was recovered after an extensive amount of creep had damaged the vessel wall. The computed lower head temperatures were found to be significantly lower (by more than 300 K in this particular example) with recovery relative to the case without recovery. This clearly demonstrates the potential for in-vessel cooling of the reactor vessel without a need to externally submerge the lower head should such a severe accident occur as core melting and relocation. (orig.)

Welding in repair of nuclear reactor pressure vessels

International Nuclear Information System (INIS)

Pilous, V.; Kovarik, R.

1987-01-01

Specific welding conditions are described in repair of the pressure vessels of nuclear reactors in operation and the effect is pointed out to of neutrons on changes in steel properties. Some of the special regulations are discussed to be observed in welding jobs. The welding methods are briefly described; the half-bead method is most frequently used. It is stressed that the defect must first be identified using a nondestructive method and the stages must be defined of the welding repair of the pressure vessel. (J.B.). 4 figs., 1 tab., 16 refs

Finite difference program for calculating hydride bed wall temperature profiles

International Nuclear Information System (INIS)

Klein, J.E.

1992-01-01

A QuickBASIC finite difference program was written for calculating one dimensional temperature profiles in up to two media with flat, cylindrical, or spherical geometries. The development of the program was motivated by the need to calculate maximum temperature differences across the walls of the Tritium metal hydrides beds for thermal fatigue analysis. The purpose of this report is to document the equations and the computer program used to calculate transient wall temperatures in stainless steel hydride vessels. The development of the computer code was motivated by the need to calculate maximum temperature differences across the walls of the hydrides beds in the Tritium Facility for thermal fatigue analysis

Mechanical behaviour of the reactor vessel support of a pressurized water reactor: tests and analysis

International Nuclear Information System (INIS)

Bolvin, M.; L'huby, Y.; Quillico, J.J.; Humbert, J.M.; Thomas, J.P.; Hugenschmitt, R.

1985-08-01

The PWR reactor vessel is supported by a steel ring laying on the reactor pit. This support has to ensure a good behaviour of the vessel in the event of accidental conditions (earthquake and pipe rupture). A new evolution of the evaluation methods of the applied forces has shown a significant increase in the design loads used until now. In order to take into account these new forces, we carried out a test on a representative mock-up of the vessel support (scale 1/6). This test was performed by CEA, EDF and FRAMATOME. Several static equivalent forces were applied on the experimental mock-up. Displacements and strains were simultaneously recorded. The results of the test have enabled to justify the design of the pit and the ring, to show up a wide safety margin until the collapse of the structures and to check our hypothesis about the transmission of the forces between the ring and the pit

Survey on Cooled-Vessel Designs in High Temperature Gas-Cooled Reactors

International Nuclear Information System (INIS)

Kim, Min-Hwan; Lee, Won-Jae

2006-01-01

The core outlet temperature of the coolant in the high temperature gas-cooled reactors (HTGR) has been increased to improve the overall efficiency of their electricity generation by using the Brayton cycle or their nuclear hydrogen production by using thermo-chemical processes. The increase of the outlet temperature accompanies an increase of the coolant inlet temperature. A high coolant inlet temperature results in an increase of the reactor pressure vessel (RPV) operation temperature. The conventional steels, proven vessel material in light water reactors, cannot be used as materials for the RPV in the elevated temperatures which necessitate its design to account for the creep effects. Some ferritic or martensitic steels like 2 1/4Cr-1Mo and 9Cr-1Mo-V are very well established creep resistant materials for a temperature range of 400 to 550 C. Although these materials have been used in a chemical plant, there is limited experience with using these materials in nuclear reactors. Even though the 2 1/4Cr-1Mo steel was used to manufacture the RPV for HTR-10 of Japan Atomic Energy Agency(JAEA), a large RPV has not been manufactured by using this material or 9Cr-1Mo-V steel. Due to not only its difficulties in manufacturing but also its high cost, the JAEA determined that they would exclude these materials from the GTHTR design. For the above reasons, KAERI has been considering a cooled-vessel design as an option for the RPV design of a NHDD plant (Nuclear Hydrogen Development and Demonstration). In this study, we surveyed several HTGRs, which adopt the cooled-vessel concept for their RPV design, and discussed their design characteristics. The survey results in design considerations for the NHDD cooled-vessel design

Remote controlled ultrasonic pre-service and in-service inspections of reactor pressure vessels

International Nuclear Information System (INIS)

Mueller, G.

1990-01-01

The first mechanised in-service inspection of the reactor pressure vessel on unit one of Eskom's Koeberg nuclear power station has been carried out. Since 1968 a whole range of manipulators to carry out remote controlled ultrasonic inspections of nuclear power station equipment has been developed. The inspection of a reactor pressure vessel using a central mast manipulator is described. 3 figs., 1 ill

Ageing study of Cirus reactor vessel expansion bellow

International Nuclear Information System (INIS)

Ramana, W.V.; Dutta, B.K.; Kushwaha, H.S.; Sahu, A.K.; Bhatnagar, A.; Pant, R.C.

1994-01-01

Expansion bellow of Cirus reactor vessel is a comparatively weak component which is joined to top tube sheet and shell by helium tight lap weld. This has been subjected to thermal stress caused by high temperature during reactor operation and thermal shock due to trip or shutdown. Therefore a finite element analysis was carried out to assess thermal stresses and fatigue life of the component. It was found that the fluctuating stress in the bellow is far less than its endurance limit. (author). 2 tabs., 3 figs

In-service inspection program for the NCS-80 reactor pressure vessel

International Nuclear Information System (INIS)

Scharge, J.; Wehowsky, P.; Zeibig, H.

1978-01-01

The in-service inspection program of reactor pressure vessels is mainly based on the ultra-sonic method, visual checking of inner and outer surfaces as well as pressure and leak tests. The test procedure require a design of the pressure vessel suitable for the test methods and the possibility to remove the pressure vessel internals. For the outside inspection a gap of sufficient width is mandatory. The present status of the ultra-sonic method and of the inner and outer manipulators affords to conduct the in-service inspection program in form of automatic checkings. The in-service inspection program for NCS-80, the Nuclear Container-Ship design of 80,000 shp, is integrated in the refueling periods due to the request for a high availability of the ship and reactor plant

Latest developments in prestressed concrete vessels for gas-cooled reactors

International Nuclear Information System (INIS)

Ople, F.S. Jr.

1979-01-01

This paper is an update of the design development of prestressed concrete vessels, commonly referred to as 'PCRVs' starting with the first single-cavity PCRV for the Fort St. Vrain Nuclear Generating Station to the latest multi-cavity PCRV configurations being utilized as the primary reactor vessels for both the High Temperature Gas-Cooled Reactor (HTGR) and the Gas-Cooled Fast Breeder Reactor (GCFR) in the U.S.A. The complexity of PCRV design varies not only due to the type of vessel configuration (single versus multi-cavity) but also on the application to the specific type of reactor concept. PCRV technology as applied to the Steam Cycle HTGR is fairly well established; however, some significant technical complexities are associated with PCRV design for the Gas Turbine HTGR and the GCFR. For the Gas Turbine HTGR, for instance, the fluid dynamics of the turbo-machinery cause multi-pressure conditions to exist in various portions of the power conversion loops during operation. This condition complicates the design approach and the proof test specification for the PCRV. The geometric configuration of the multi-cavity PCRV is also more complex due to the introduction of large horizontal cylindrical cavities (housing the turbo/machines for the Gas Turbine HTGR and circulators for the GCFR) in addition to the vertical cylindrical cavities for the core and heat exchangers. Because of this complex geometry, it becomes difficult to achieve an optimum prestressing arrangement for the PCRV. Other novel features of the multi-cavity PCRV resulting from the continuing design optimization effort are the incorporation of an asymmetric (offset core) configuration and the use of large vessel cavity/penetration concrete closures directly held down by prestressing tendons for both economic and safety reasons. (orig.)

Development of advanced design features for KNGR reactor vessel and internals

Energy Technology Data Exchange (ETDEWEB)

Park, Jong Kyun; Ru, Bong; Lee, Jae Han; Lee, Hyung Yeon; Kim, Jong Bum; Ku, Kyung Heoy; Lee, Ki Young; Lee, Jun; Kim, Young In [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

1995-12-01

Developments of KNGR design require to enhance the design to implement the design requirements, such as plant life time from 40 years to 60 years, safety, performance and structure and components design. The designs used for existing nuclear power plants should be modified or improved to meet the requirements in KNGR design. The purpose of the task is to develop the Advanced Design Features (ADF) related to mechanical and structural design for KNGR reactor vessel and reactor internals. The structural integrity for the System 80+ reactor vessel, of which design life is 60 years, was reviewed. EPRI-URD, CESSAR-DC, and the present design status and characteristics of System 80+ reactor vessel were comparatively studied and the improvement of reactor vessel surveillance program was investigated. The performance and aseismic characteristics of the CE-type CEDM, which will be used in System 80+, are investigated. The driving cycles of CEDM are evaluated for the load follow operation(LFO), of which Mode K is being developed by KAERI. The position of the USNRC, EPRI, ABB-CE, and industries on the elimination of OBE are reviewed, and especially ABB-CE System 80+ FSER is reviewed in detail. For the pre-stage of the verification of the OBE elimination from the design, the review of the seismic responses, i.e.. shear forces and moments, of YGN 3/4 RI was performed and the ratio of OBE response to SSE response was analysed. The screening criteria were reviewed to evaluate the integrity against pressurized thermal shock (PTS) for RV belt-line of System 80+. The evaluation methods for fracture integrity when screening criteria are not met were reviewed. The structural characteristics of IRWST spargers of System 80+ were investigated and the effect of hydrodynamic loads on NSSS was reviewed. 18 figs., 9 tabs., 40 refs. (Author) .new.

Development of advanced design features for KNGR reactor vessel and internals

International Nuclear Information System (INIS)

Park, Jong Kyun; Ru, Bong; Lee, Jae Han; Lee, Hyung Yeon; Kim, Jong Bum; Ku, Kyung Heoy; Lee, Ki Young; Lee, Jun; Kim, Young In

1995-12-01

Developments of KNGR design require to enhance the design to implement the design requirements, such as plant life time from 40 years to 60 years, safety, performance and structure and components design. The designs used for existing nuclear power plants should be modified or improved to meet the requirements in KNGR design. The purpose of the task is to develop the Advanced Design Features (ADF) related to mechanical and structural design for KNGR reactor vessel and reactor internals. The structural integrity for the System 80+ reactor vessel, of which design life is 60 years, was reviewed. EPRI-URD, CESSAR-DC, and the present design status and characteristics of System 80+ reactor vessel were comparatively studied and the improvement of reactor vessel surveillance program was investigated. The performance and aseismic characteristics of the CE-type CEDM, which will be used in System 80+, are investigated. The driving cycles of CEDM are evaluated for the load follow operation(LFO), of which Mode K is being developed by KAERI. The position of the USNRC, EPRI, ABB-CE, and industries on the elimination of OBE are reviewed, and especially ABB-CE System 80+ FSER is reviewed in detail. For the pre-stage of the verification of the OBE elimination from the design, the review of the seismic responses, i.e.. shear forces and moments, of YGN 3/4 RI was performed and the ratio of OBE response to SSE response was analysed. The screening criteria were reviewed to evaluate the integrity against pressurized thermal shock (PTS) for RV belt-line of System 80+. The evaluation methods for fracture integrity when screening criteria are not met were reviewed. The structural characteristics of IRWST spargers of System 80+ were investigated and the effect of hydrodynamic loads on NSSS was reviewed. 18 figs., 9 tabs., 40 refs. (Author) .new

Damage by radiation in structural materials of BWR reactor vessels; Dano por radiacion en materiales estructurales de vasijas de reactores BWR

Energy Technology Data Exchange (ETDEWEB)

Robles, E.; Balcazar, M.; Alpizar, A.M.; Calderon, B.E. [Departamento de Sintesis y Caracterizacion de Materiales, Instituto Nacional de Investigaciones Nucleares, A.P. 18-1027, 11801 Mexico D.F. (Mexico)

2002-07-01

The structural materials which are manufactured the pressure vessels of the BWR reactors undergo degradation in their mechanical properties mainly due to the damage produced by the fast neutrons (E> 1 MeV) coming from the reactor core. The mechanisms of neutron damage in this type of materials are experimentally studied, through the irradiation of vessel steel in experimental reactors for a quickly ageing. Alternately the neutron damage through steel irradiation with heavy ions is simulated. In this work the first results of the damage induced by irradiation of a similar steel to the vessel of a BWR reactor are shown. The irradiation was performed with fast neutrons (E> 1 MeV, fluence of 1.45 x 10{sup 18} n/cm{sup 2}) in the TRIGA Mark III Salazar reactor and separately with Ni{sup +3} ions in a Tandetrom accelerator (E= 4.8 MeV and an ion flux rank of 0.1 to 53 ions/A{sup 2}). (Author)

Studies of fragileness in steels of vessels of BWR reactors; Estudios de fragilizacion en aceros de vasija de reactores BWR

Energy Technology Data Exchange (ETDEWEB)

Robles, E.F.; Balcazar, M.; Alpizar, A.M.; Calderon, B.E. [ININ, 52045 Ocoyoacac, Estado de Mexico (Mexico)

2003-07-01

The structural materials with those that are manufactured the pressure vessels of the BWR reactors, suffer degradation in its mechanical properties mainly to the damage taken place by the fast neutrons (E > 1 MeV) coming from the reactor core. Its are experimentally studied those mechanisms of neutron damage in this material type, by means of the irradiation of steel vessel in experimental reactors to age them quickly. Alternatively it is simulated the neutron damage by means of irradiation of steel with heavy ions. In this work those are shown first results of the damage induced by irradiation from a similar steel to the vessel of a BWR reactor. The irradiation was carried out with fast neutrons (E > 1 MeV, fluence of 1.45 x 10{sup 18} n/cm{sup 2}) in the TRIGA MARK lll reactor and separately with Ni{sup +3} ions in a Tandetrom accelerator, E = 4.8 MeV and range of the ionic flow of 0.1 to 53 iones/A{sup 2}. (Author)

Seals for sealing a pressure vessel such as a nuclear reactor vessel or the like

International Nuclear Information System (INIS)

Bruns, H.J.; Huelsermann, K.H.

1975-01-01

A description is given of seals for sealing a pressure vessel such as a nuclear reactor vessel, steam boiler vessel, or any other vessel which is desirably sealed against pressure of the type including a housing and a housing closure that present opposed vertical sealing surfaces which define the sides of a channel. The seals of the present invention comprise at least one sealing member disposed in the channel, having at least one stop face, a base portion and two shank portions extending from the base portion to form a groove-like recess. The shank portions are provided with sealing surfaces arranged to mate with the opposed vertical pressure vessel sealing surfaces. A shank-spreading wedge element also disposed in the channel has at least one stop face and is engaged in the groove-like recess with the sealing member and wedge element stop face adjacent to each other

Flaw density examinations of a clad boiling water reactor pressure vessel segment

International Nuclear Information System (INIS)

Cook, K.V.; McClung, R.W.

1986-01-01

Flaw density is the greatest uncertainty involved in probabilistic analyses of reactor pressure vessel failure. As part of the Heavy-Section Steel Technology (HSST) Program, studies have been conducted to determine flaw density in a section of reactor pressure vessel cut from the Hope Creek Unit 2 vessel [nominally 0.7 by 3 m (2 by 10 ft)]. This section (removed from the scrapped vessel that was never in service) was evaluated nondestructively to determine the as-fabricated status. We had four primary objectives: (1) evaluate longitudinal and girth welds for flaws with manual ultrasonics, (2) evaluate the zone under the nominal 6.3-mm (0.25-in.) clad for cracking (again with manual ultrasonics), (3) evaluate the cladding for cracks with a high-sensitivity fluorescent penetrant method, and (4) determine the source of indications detected

Titanium tritide radioisotope heat source development: palladium-coated titanium hydriding kinetics and tritium loading tests

International Nuclear Information System (INIS)

Van Blarigan, Peter; Shugard, Andrew D.; Walters, R. Tom

2012-01-01

We have found that a 180 nm palladium coating enables titanium to be loaded with hydrogen isotopes without the typical 400-500 C vacuum activation step. The hydriding kinetics of Pd coated Ti can be described by the Mintz-Bloch adherent film model, where the rate of hydrogen absorption is controlled by diffusion through an adherent metal-hydride layer. Hydriding rate constants of Pd coated and vacuum activated Ti were found to be very similar. In addition, deuterium/tritium loading experiments were done on stacks of Pd coated Ti foil in a representative-size radioisotope heat source vessel. The experiments demonstrated that such a vessel could be loaded completely, at temperatures below 300 C, in less than 10 hours, using existing department-of-energy tritium handling infrastructure.

Study on external reactor vessel cooling capacity for advanced large size PWR

International Nuclear Information System (INIS)

Jin Di; Liu Xiaojing; Cheng Xu; Li Fei

2014-01-01

External reactor vessel cooling (ERVC) is widely adopted as a part of in- vessel retention (IVR) in severe accident management strategies. In this paper, some flow parameters and boundary conditions, eg., inlet and outlet area, water inlet temperature, heating power of the lower head, the annular gap size at the position of the lower head and flooding water level, were considered to qualitatively study the effect of them on natural circulation capacity of the external reactor vessel cooling for an advanced large size PWR by using RELAP5 code. And the calculation results provide some basis of analysis for the structure design and the following transient response behavior of the system. (authors)

New paradigm for prediction of radiation life-time of reactor pressure vessel

International Nuclear Information System (INIS)

Kotrechko, S.A.; Meshkov, Yu.Ya.; Neklyudov, I.M.; Revka, V.N.

2011-01-01

New paradigm for prediction of radiation life-time of reactor pressure vessel is presented. Equation for limiting state of reactor pressure vessel wall with crack-like defect is obtained. It is exhibited that the value of critical fluence Φ c may be determined not by shift of critical temperature of fracture of surveillance specimen, which is indirect characteristic, but by direct method, namely, by the condition of initiation of brittle fracture of irradiated metal ahead of a crack in RPV wall. Within the framework of engineering version of LA to fracture the technique for Φ c ascertainment is developed. Prediction of Φ c for WWER pressure vessels demonstrates potentialities of this technique.

Reactor containment vessel

International Nuclear Information System (INIS)

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

1981-01-01

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

Proceedings of the twenty-fourth water reactor safety information meeting. Volume 2: Reactor pressure vessel embrittlement and thermal annealing; Reactor vessel lower head integrity; Evaluation and projection of steam generator tube condition and integrity

Energy Technology Data Exchange (ETDEWEB)

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

1997-02-01

This three-volume report contains papers presented at the Twenty-Fourth Water Reactor Safety Information Meeting held at the Bethesda Marriott Hotel, Bethesda, Maryland, October 21--23, 1996. The papers are printed in the order of their presentation in each session and describe progress and results of programs in nuclear safety research conducted in this country and abroad. Foreign participation in the meeting included papers presented by researchers from Czech Republic, Finland, France, Japan, Norway, Russia and United Kingdom. This volume is divided into the following sections: reactor pressure vessel embrittlement and thermal annealing; reactor vessel lower head integrity; and evaluation and projection of steam generator tube condition and integrity. Selected papers are indexed separately for inclusion in the Energy Science and Technology Database.

Proceedings of the twenty-fourth water reactor safety information meeting. Volume 2: Reactor pressure vessel embrittlement and thermal annealing; Reactor vessel lower head integrity; Evaluation and projection of steam generator tube condition and integrity

International Nuclear Information System (INIS)

Monteleone, S.

1997-02-01

This three-volume report contains papers presented at the Twenty-Fourth Water Reactor Safety Information Meeting held at the Bethesda Marriott Hotel, Bethesda, Maryland, October 21--23, 1996. The papers are printed in the order of their presentation in each session and describe progress and results of programs in nuclear safety research conducted in this country and abroad. Foreign participation in the meeting included papers presented by researchers from Czech Republic, Finland, France, Japan, Norway, Russia and United Kingdom. This volume is divided into the following sections: reactor pressure vessel embrittlement and thermal annealing; reactor vessel lower head integrity; and evaluation and projection of steam generator tube condition and integrity. Selected papers are indexed separately for inclusion in the Energy Science and Technology Database

VISA-2, Reactor Vessel Failure Probability Under Thermal Shock

International Nuclear Information System (INIS)

Simonen, F.; Johnson, K.

1992-01-01

1 - Description of program or function: VISA2 (Vessel Integrity Simulation Analysis) was developed to estimate the failure probability of nuclear reactor pressure vessels under pressurized thermal shock conditions. The deterministic portion of the code performs heat transfer, stress, and fracture mechanics calculations for a vessel subjected to a user-specified temperature and pressure transient. The probabilistic analysis performs a Monte Carlo simulation to estimate the probability of vessel failure. Parameters such as initial crack size and position, copper and nickel content, fluence, and the fracture toughness values for crack initiation and arrest are treated as random variables. Linear elastic fracture mechanics methods are used to model crack initiation and growth. This includes cladding effects in the heat transfer, stress, and fracture mechanics calculations. The simulation procedure treats an entire vessel and recognizes that more than one flaw can exist in a given vessel. The flaw model allows random positioning of the flaw within the vessel wall thickness, and the user can specify either flaw length or length-to-depth aspect ratio for crack initiation and arrest predictions. The flaw size distribution can be adjust on the basis of different inservice inspection techniques and inspection conditions. The toughness simulation model includes a menu of alternative equations for predicting the shift in the reference temperature of the nil-ductility transition. 2 - Method of solution: The solution method uses closed form equations for temperatures, stresses, and stress intensity factors. A polynomial fitting procedure approximates the specified pressure and temperature transient. Failure probabilities are calculated by a Monte Carlo simulation. 3 - Restrictions on the complexity of the problem: Maxima of 30 welds. VISA2 models only the belt-line (cylindrical) region of a reactor vessel. The stresses are a function of the radial (through-wall) coordinate only

The Use of Prestressed Concrete Vessels in the French Power Reactor Programme

International Nuclear Information System (INIS)

Conte, F.; Dambrine, C.; Gaussot, D.

1963-01-01

This paper deals with the use of pre-stressed concrete for the G2 and G3 reactors at Marcoule and for the EDF3 reactor now under construction at Chinon. The first two reactors have been operating at power since 1959 and 1960 respectively. Messrs. Conte and Dambrine discuss the problems that arose during construction of the vessels for G2 and G3 and also deal with the experience gained in operation - experience which suggests that they are extremely safe- Work on the EDF3 vessel, begun at Chinon in the second half of 1961, is still under way and should be finished towards the end of 1963. Mr. Gaussot discusses the reasons for choosing this type of vessel, the results of calculations and mock-up tests, and the problems presented by the construction itself. A number of studies have been devoted to the future prospects of prestressed concrete structures for reactors. It would seem that working pressures could be increased, if desired, and, in any case, that dimensions could be considerably enlarged, thus offering the chance of integral-type solutions. (author) [fr

Evaluation of the integrity of reactor vessels designed to ASME Code, Sections I and/or VIII

International Nuclear Information System (INIS)

Hoge, K.G.

1976-01-01

A documented review of nuclear reactor pressure vessels designed to ASME Code, Sections I and/or VIII is made. The review is primarily concerned with the design specifications and quality assurance programs utilized for the reactor vessel construction and the status of power plant material surveillance programs, pressure-temperature operating limits, and inservice inspection programs. The following ten reactor vessels for light-water power reactors are covered in the report: Indian Point Unit No. 1, Dresden Unit No. 1, Yankee Rowe, Humboldt Bay Unit No. 3, Big Rock Point, San Onofre Unit No. 1, Connecticut Yankee, Oyster Creek, Nine Mile Point Unit No. 1, and La Crosse

Heat treatment device for extending the life of a pressure vessel, particularly a reactor pressure vessel

International Nuclear Information System (INIS)

Krauss, P.; Mueller, E.; Poerner, H.; Weber, R.

1979-01-01

A support body in the form of an insulating cylinder is tightly sealed by connected surfaces at its outer circumference to the inner wall of the pressure vessel. It forms an annular heating space. The heat treatment or tempering of the pressure vessel takes place with the reactor space empty and screened from the outside by ceiling bolts. Heating gas or an induction winding can be used as the means of heating. (DG) [de

Cooling of pressurized water nuclear reactor vessels

International Nuclear Information System (INIS)

Curet, H.D.

1978-01-01

The improvement of pressurized water nuclear reactor vessels comprising flow dividers providing separate and distinct passages for the flow of core coolant water from each coolant water inlet, the flow dividers being vertically disposed in the annular flow areas provided by the walls of the vessel, the thermal shield (if present), and the core barrel is described. In the event of rupture of one of the coolant water inlet lines, water, especially emergency core coolant water, in the intact lines is thus prevented from by-passing the core by circumferential flow around the outermost surface of the core barrel and is instead directed so as to flow vertically downward through the annulus area between the vessel wall and the core barrel in a more normal manner to increase the probability of cooling of the core by the available cooling water in the lower plenum, thus preventing or delaying thermal damage to the core, and providing time for other appropriate remedial or damage preventing action by the operator

Seismic proving test of PWR reactor containment vessel

International Nuclear Information System (INIS)

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

1987-01-01

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

Standard Guide for In-Service Annealing of Light-Water Moderated Nuclear Reactor Vessels

CERN Document Server

American Society for Testing and Materials. Philadelphia

2003-01-01

1.1 This guide covers the general procedures to be considered for conducting an in-service thermal anneal of a light-water moderated nuclear reactor vessel and demonstrating the effectiveness of the procedure. The purpose of this in-service annealing (heat treatment) is to improve the mechanical properties, especially fracture toughness, of the reactor vessel materials previously degraded by neutron embrittlement. The improvement in mechanical properties generally is assessed using Charpy V-notch impact test results, or alternatively, fracture toughness test results or inferred toughness property changes from tensile, hardness, indentation, or other miniature specimen testing (1). 1.2 This guide is designed to accommodate the variable response of reactor-vessel materials in post-irradiation annealing at various temperatures and different time periods. Certain inherent limiting factors must be considered in developing an annealing procedure. These factors include system-design limitations; physical constrain...

A review of formulas for predicting irradiation embrittlement of reactors vessel materials

International Nuclear Information System (INIS)

Petrequin, P.

1995-01-01

Formulas developed in different countries for predicting irradiation embrittlement of reactors vessel materials are presented. Results of predictions were compared with different data sets, from surveillance programmes or studies in test reactors, with different residual elements contents. Figs

Sub-critical crack growth and clad integrity in a PWR reactor pressure vessel

International Nuclear Information System (INIS)

Tice, D.R.; Foreman, A.J.E.; Sharples, J.K.

1987-10-01

The possibility of in-service growth of sub-critical defects in a PWR reactor pressure vessel to a critical size which could result in vessel failure was addressed in both the 1976 and 1982 reports of the Light Water Reactor Study Group (LWRSG), under the Chairmanship of Dr W Marshall (now Lord Marshall). An addendum to this report was published by UKAEA in April 1987. The section of the addendum dealing with subcritical crack growth and the related issue of integrity of the stainless steel cladding on the inner vessel surface is reproduced in this report. This section of the LWRSG addendum provides a review of the current status of fatigue crack growth and environmentally assisted cracking research for pressure vessel steels in light water reactor environments, as well as a review of developments in crack growth assessment methods. The review concludes that the alternative assessment procedures now being developed give a more realistic prediction of in service crack growth than the ASME Section XI Appendix A fatigue crack growth curves. (author)

Influence of temperature on δ-hydride habit plane in α-Zirconium

International Nuclear Information System (INIS)

Singh, R. N.; Stahle, P.; Banerjee, S.; Ristmanaa, Matti; Sauramd, K.

2008-01-01

Dilute Zr-alloy with hcp α-Zr as major phase is used as pressure boundary for hot coolant in CANDU, PHWR and RBMK reactors. Hydrogen / deuterium ingress during service makes the pressure boundary components like pressure tubes of the aforementioned reactors susceptible to hydride embrittlement. Hydride acquires plate shaped morphology and the broad face of the hydride plate coincides with certain crystallographic plane of α-Zr crystal, which is called habit plane. Hydride plate oriented normal to tensile stress significantly increases the degree of embrittlement. Thus key to mitigating the damage due to hydride embrittlement is to avoid the formation of hydride plates normal to tensile stress. Two different theoretical approaches are used to determine the habit plane of precipitates viz., geometrical and solid mechanics. For the geometrical approach invariant plane and invariant-line criteria have been applied successfully and for the solid mechanics approach strain energy minimization criteria have been used successfully. Solid mechanics approach using strain energy computed by FEM technique has been applied to hydride precipitation in Zr-alloys, but the emphasis has been to understand the solvus hysteresis. The objective of the present investigation is to predict the habit plane of δ-hydride precipitating in α-Zr at 25, 300, 400 and 450 .deg. C. using strain energy minimization technique. The δ-hydride phase is modeled to undergo isotropic elastic and plastic deformation. The α-Zr phase was modeled to undergo transverse isotropic elastic deformation. Both isotropic plastic and transverse isotropic plastic deformations of α-Zr were considered. Further, both perfect and linear work-hardening plastic behaviors were considered. Accommodation strain energy of δ-hydrides forming in α-Zr crystal was computed using initial strain method as a function of hydride nuclei orientation. Hydride was modeled as disk with circular edge. The simulation was carried out

Performance study of a hydrogen powered metal hydride actuator

International Nuclear Information System (INIS)

Bhuiya, Md Mainul Hossain; Kim, Kwang J

2016-01-01

A thermally driven hydrogen powered actuator integrating metal hydride hydrogen storage reactor, which is compact, noiseless, and able to generate smooth actuation, is presented in this article. To test the plausibility of a thermally driven actuator, a conventional piston type actuator was integrated with LaNi 5 based hydrogen storage system. Copper encapsulation followed by compaction of particles into pellets, were adopted to improve overall thermal conductivity of the reactor. The operation of the actuator was thoroughly investigated for an array of operating temperature ranges. Temperature swing of the hydride reactor triggering smooth and noiseless actuation over several operating temperature ranges were monitored for quantification of actuator efficiency. Overall, the actuator generated smooth and consistent strokes during repeated cycles of operation. The efficiency of the actuator was found to be as high as 13.36% for operating a temperature range of 20 °C–50 °C. Stress–strain characteristics, actuation hysteresis etc were studied experimentally. Comparison of stress–strain characteristics of the proposed actuator with traditional actuators, artificial muscles and so on was made. The study suggests that design modification and use of high pressure hydride may enhance the performance and broaden the application horizon of the proposed actuator in future. (paper)

Recent evaluation of 'wet' thermal annealing to resolve reactor pressure vessel embrittlement

International Nuclear Information System (INIS)

Server, W.L.; Biemiller, E.C.

1993-01-01

Prior to the decision to close the Yankee Rowe plant in 1992, a great deal of effort was expended in trying to resolve the degree of neutron embrittlement that the reactor pressure vessel had experienced after 30 years of operation. One mitigative measure that was examined in detail was the possibility of performing a relatively low temperature thermal anneal (at approximately 650 deg. F) to partially restore the original design level of mechanical properties of the reactor pressure vessel beltline region which were lost due to the neutron radiation exposure. This low temperature anneal was to involve heating of the primary coolant water using pump heat in a similar manner as that used to anneal the Belgian BR-3 reactor pressure vessel in the early 1980s. This 'wet' anneal was successful in recovering mechanical properties for the BR-3 vessel, but the extent of the recovery, as well as the rate of re-embrittlement after the anneal, were issues that were difficult to quantify since the exact reactor pressure vessel steels were not available for experimental verification. For the case of Yankee Rowe, material was available from past surveillance programs for at least one of the materials in the vessel, as well as materials obtained from various sources which could act as bounding surrogates. An irradiation /annealing/reirradiation program was developed to better quantify the degree of recovery and re-embrittlement for these materials, but this program was halted before significant test results were obtained. Prior to the initiation of the testing program, a review of past annealing data was performed and the data were scrutinized for direct relevance to the annealing response of the Yankee Rowe vessel. This paper discusses the results derived from this review. The results from the critical review of the past annealing data indicated that a 'wet' anneal of the Yankee Rowe vessel may have been successful in reducing the degree of embrittlement to the point that the

The 4th surveillance testing for Kori unit 3 reactor vessel materials

Energy Technology Data Exchange (ETDEWEB)

Chang, Kee Ok; Kim, Byoung Chul; Lee, Sam Lai; Choi, Kwun Jae; Gong, Un Sik; Chang, Jong Hwa; Joo, Yong Sun; Ahn, Sang Bok; Hong, Joon Hwa [Korea Atomic Energy Research Institute, Taejeon (Korea)

2000-10-01

Surveillance testing for reactor vessel materials is performed in order to evaluate the irradiation embrittlement due to neutrons during operation and set the condition of safe operation of nuclear reactor. The 4th surveillance testing was performed completely by Korea Atomic Energy Research Institute at Taejeon after the capsule was transported from Kori site including its removal from reactor. Fast neutron fluences for capsules were calculated and various testing including mechanical and chemistry analysis were performed in order to evaluate the integrity of Kori unit 3 reactor vessel during the operation until life time. The evaluation results are as follows; Fast neutron fluences for capsules U, V, X and W are 4.983E+18, 1.641E+19, 3.158E+19, and 4.469E+19n/cm{sup 2}, respectively. The bias factor, the ratio of calculation/measurement, was 0.840 for the 1st through 4th testing and the calculational uncertainty, 12% satisfied the requirement of USNRC Reg.Guide DG-1053, 20%. The best estimated neutron fluence for reactor vessel inside surface was 1.362E+19n/cm{sup 2} based on the end of 12th fuel cycle and it was predicted that the fluences of vessel inside surface at 32, 40, 48 and 56EFPY would reach 3.481E+19, 4.209E+19, 5.144E+19 and 5.974E+19n/cm{sup 2} based on the current calculation. The result through this analysis for Kori unit 3 showed that there would be no problem for the pressurized thermal shock(PTS) during the operation until design life. 48 refs., 35 figs., 41 tabs. (Author)

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

International Nuclear Information System (INIS)

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

2004-01-01

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

In-service inspection of nuclear reactor vessels and steam generators. Results and evolution of the technics

International Nuclear Information System (INIS)

Rapin, Michel; Saglio, Robert.

1978-01-01

Methods and original technics have been developed by the CEA for inspection of the primary coolant circuit of PWR. Multifrequency Eddy currents for inspection of steam generators tubes gudgeons and bolts; focussed ultrasonics to test all the welds of the reactor vessel and its cover of mixed welds of tanks and steam generators, pressurizer welds and gudgeons from the inside; gamma radiography of vessel mixed welds, televisual examination of the stainless steel lining of the reactor vessel and its cover. Use of these technics is made with specific automatic machines designed either for inspection of steam generator tubes or for complete inspection of the vessel. Several reactors were inspected with these devices [fr

Fracture toughness requirements of reactor vessel material in evaluation of the safety analysis report of nuclear power plants

International Nuclear Information System (INIS)

Widia Lastana Istanto

2011-01-01

Fracture toughness requirements of reactor vessel material that must be met by applicants for nuclear power plants construction permit has been investigated in this paper. The fracture toughness should be described in the Safety Analysis Reports (SARs) document that will be evaluated by the Nuclear Energy Regulatory Agency (BAPETEN). Because BAPETEN does not have a regulations or standards/codes regarding the material used for the reactor vessel, especially in the fracture toughness requirements, then the acceptance criteria that applied to evaluate the fracture toughness of reactor vessel material refers to the regulations/provisions from the countries that have been experienced in the operation of nuclear power plants, such as from the United States, Japan and Korea. Regulations and standards used are 10 CFR Part 50, ASME and ASTM. Fracture toughness of reactor vessel materials are evaluated to ensure compliance of the requirements and provisions of the Regulatory Body and the applicable standards, such as ASME or ASTM, in order to assure a reliability and integrity of the reactor vessels as well as providing an adequate safety margin during the operation, testing, maintenance, and postulated accident conditions over the reactor vessel lifetime. (author)

Effectiveness of External Reactor Vessel Cooling (ERVC) strategy for APR1400 and issues of phenomenological uncertainties

International Nuclear Information System (INIS)

Oh, S.J.; Kim, H.T.

2007-01-01

The APR1400(Advanced Power Reactor 1400) is an evolutionary advanced light water reactor with rated thermal power of 4000 MWt. For APR1400, External Reactor Vessel Cooling (ERVC) is adopted as a primary severe accident management strategy for in-vessel retention (IVR) of corium. The ERVC is a method of IVR by submerging the reactor vessel exterior. At the early stage of the APR1400 design, only ex-vessel cooling, cooling of the core melt outside the vessel after vessel is breached, is considered based on the EPRI Utility Requirement Document for Evolutionary LWR. However, based on the progress in implementation of Severe Accident Management Guidance (SAMG) for operating plants, as well as the research findings related to ERVC, ERVC strategy is adopted as a part of key severe accident management strategies. To improve its success, the strategy is reviewed and we implemented necessary design arrangement to increase its usefulness in managing the severe accident. In this paper, we examine the evolution of ERVC concept and its implementation in APR1400. Then, we review possible approach, including Risk-Oriented Accident Analysis Methodology (ROAAM), to evaluate the effectiveness of the strategy. (authors)

Aging impact on the safety and operability of nuclear reactor pressure vessels

International Nuclear Information System (INIS)

Pennell, W.E.

1992-01-01

Irradiation embrittlement causes a loss of reactor vessel material fracture toughness as nuclear plants age. Fracture mechanics based regulatory requirements limit the permissible level of irradiation embrittlement such that essential fracture prevention margins are maintained throughout the plant operating life. This paper reviews the regulatory requirements and the underlying fracture mechanics technology. Issues identified with that technology are identified and research programs implemented to resolve the issues are described. Where possible, an assessment is given of the anticipated impact on the research program output will have on the reactor vessel fracture-margin assessment process

Prediction of thermal margin for external cooling of reactor vessel lower head during a severe accident

International Nuclear Information System (INIS)

Yoon, Ho Jun; Suh, Kune Y.

1998-01-01

In the TMI-2 accident, approximately nineteen (19) tons of molten core material drained into the lower plenum. One of the major findings from the TMI-2 Vessel Investigation Project was that one part of the reactor lower head wall estimated to have attained a temperature of 1100 .deg. C for about 30 minutes has seemingly experienced a comparatively rapid cooldown with no major threat to the vessel integrity. In this regard, recent empirical and analytical studies have shifted interests to such in-vessel retention designs or strategies as reactor cavity flooding, in-vessel flooding and engineered gap cooling of the vessel. Accurate thermohydrodynamic and creep deformation modeling and rupture prediction are the key to the success in developing practically useful in-vessel accident management strategies. As an advanced in-vessel design concept, the COrium Attak Syndrome Immunization Structures (COASIS) are being developed as prospective in-vessel retention devices for a next-generation LWR in concert with existing ex-vessel management measures. Both the engineered gap structures in -vessel (COASISI) and ex-vessel (COASISO) were demonstrated to maintain effective heat transfer geometry during molten core debris attack when applied to the TMI-2 and the Korean Standard Nuclear Power Plant (KSNPP) reactors. The likelihood of lower head creep rupture during a severe accident is found to be significantly suppressed by the COASIS options. In studying the in-vessel severe accident phenomena, one of the main goals is to verify the cooling mechanism in the reactor vessel lower plenum and thereby to prevent the vessel failure from thermal attack by the molten debris. This paper presents the first-principle calculation results for the thermal margin for the case of external cooling of the reactor vessel lower head. Adopting the method presented by F.B. Cheung, et al., we calculated the departure from nucleate boiling ratio (DNBR) for the three cases of pool boiling, flow boiling

Improvement of methods to evaluate brittle failure resistance of the WWER reactor pressure vessels

Energy Technology Data Exchange (ETDEWEB)

Popov, A A; Parshutin, E V [Engineering Center of Nuclear Equipment Strength, Research and Development Inst. of Power Engineering, Moscow (Russian Federation); Rogov, M F; Dragunov, U G [Experimenter` s and Designer` s Office ` ` Hydropress` ` (Russian Federation)

1997-09-01

At the next 10 years a number of Russian WWER nuclear power plants will complete its design lifetime. Normative methods to evaluate brittle failure resistance of the reactor pressure vessels used in Russia have been intended for design stage. The evaluation of reactor pressure vessel lifetime in operation stage demands to create new methods of calculation and new methods for experimental evaluation of brittle failure resistance degradation. The main objective of the study in this type of reactor is weldment number 4. In this report an analysis is made of methods to determine critical temperature of reactor materials including the results of instrumented Charpy testing. 12 figs.

Emergency retraction mechanism for the manipulator arm of a nuclear reactor vessel inspection apparatus

International Nuclear Information System (INIS)

1980-01-01

Nuclear reactor vessels are made using numerous welds. These have to be inspected, often using ultrasonic transducers mounted on a manipulator arm. This invention seeks to solve the problem of retracting the manipulator arm should an emergency occur while it is fully extended, particularly within one of the reactor vessel nozzles. Of specific concern is the situation where power fails with the manipulator arm so extended. Details are given of an emergency retraction mechanism for use in reactor vessel inspection apparatus. A manual retraction means is used; the manipulator arm is slidably mounted within a frame. This comprises a member mounted on the arm for looping engagement by a cable, the cable being fixed at one end of the arm frame and engaging the member, and a clamp for detachably securing the cable at its other end to the arm frame at a point which is accessible from above the vessel. (U.K.)

Neutron irradiation embrittlement of reactor pressure vessel steels

International Nuclear Information System (INIS)

Steele, L.E.

1975-01-01

The reliability of nuclear power plants depends on the proper functioning of complex components over the whole life on the plant. Particular concern for reliability is directed to the primary pressure boundary. This report focuses on the portion of the primary system exposed to and significantly affected by neutron radiation. Experimental evidence from research programmes and from reactor surveillance programmes has indicated radiation embrittlement of a magnitude sufficient to raise doubts about reactor pressure vessel integrity. The crucial nature of the primary vessel function heightens the need to be alert to this problem, to which, fortunately, there are positive aspects: for example, steels have been developed which are relatively immune to radiation embrittlement. Further, awareness of such embrittlement has led to designs which can accomodate this factor. The nature of nuclear reactors, of the steels used in their construction, and of the procedures for interpreting embrittlement and minimizing the effects are reviewed with reference to the reactors that are expected to play a major role in electric power production from now to about the turn of the century. The report is intended as a manual or guidebook; the aim has been to make each chapter or major sub-division sufficiently comprehensive and self-contained for it to be understood and read independently of the rest of the book. At the same time, it is hoped that the whole is unified enough to make a complete reading useful and interesting to the several classes of reader that are involved with only specific aspects of the topic

Regulatory Assessment Technologies for Aging of Reactor Vessel Internals

Energy Technology Data Exchange (ETDEWEB)

Jhung, Myung Jo; Park, Jeong Soon; Ko, Hanok [Korea Institute of Nuclear Safety, Daejeon (Korea, Republic of)

2013-10-15

In order to develop the audit calculation system, it is required to develop crack evaluation, seismic analysis and thermal-hydraulic analysis techniques for RVIs so that integrity of RVIs under the aging environment can be evaluated and be assured. In addition, regulatory requirements including safety review and inspection guides should be developed in order to assure the quality and uniformity of safety reviews and inspections regarding aging assessment and management of RVIs. Since Reactor Vessel Internals (RVIs) are installed within the reactor pressure vessel and surround the fuel assemblies, some of them are exposed to the environment such as high neutron irradiation, high temperature and reactor coolant flow. Those environmental factors can cause damage to RVIs including cracks, loss of material, fatigue, loss of fracture toughness and change of dimension as the operation time of nuclear power plants (NPPs) increases. For long-term operation more than 40 years, aging management of RVIs is important. The final objectives of this study are to establish the audit calculation system for RVIs and to develop regulatory requirements for aging assessment and management of RVIs considering their operating conditions, materials, and possible aging mechanisms.

Imperfection detection probability at ultrasonic testing of reactor vessels

International Nuclear Information System (INIS)

Kazinczy, F. de; Koernvik, L.Aa.

1980-02-01

The report is a lecture given at a symposium organized by the Swedish nuclear power inspectorate on February 1980. Equipments, calibration and testing procedures are reported. The estimation of defect detection probability for ultrasonic tests and the reliability of literature data are discussed. Practical testing of reactor vessels and welded joints are described. Swedish test procedures are compared with other countries. Series of test data for welded joints of the OKG-2 reactor are presented. Future recommendations for testing procedures are made. (GBn)

Probabilistic fracture mechanics analysis for the life extension estimate of the high flux isotope reactor vessel

International Nuclear Information System (INIS)

Chang, S.J.

1997-01-01

The state of the vessel steel embrittlement as a result of neutron irradiation can be measured by its increase in the nil ductility temperature (NDT). This temperature is sometimes referred to as the brittle-ductile transition temperature (DBT) for fracture. The life extension of the High Flux Isotope Reactor (HFIR) vessel is calculated by using the method of fracture mechanics. A new method of fracture probability calculation is presented in this paper. The fracture probability as a result of the hydrostatic pressure test (hydrotest) is used to determine the life of the vessel. The hydrotest is performed in order to determine a safe vessel static pressure. It is then followed by using fracture mechanics to project the safe reactor operation time from the time of the satisfactory hydrostatic test. The life extension calculation provides the following information on the remaining life of the reactor as a function of the NDT increase: (1) the life of the vessel is determined by the probability of vessel fracture as a result of hydrotest at several hydrotest pressures and vessel embrittlement conditions, (2) the hydrotest time interval vs the NDT increase rate, and (3) the hydrotest pressure vs the NDT increase rate. It is understood that the use of a complete range of uncertainties of the NDT increase is equivalent to the entire range of radiation damage that can be experienced by the vessel steel. From the numerical values for the probabilities of the vessel fracture as a result of hydrotest, it is estimated that the reactor vessel life can be extended up to 50 EFPY (100 MW) with the minimum vessel operating temperature equal to 85 degrees F

Development of automatic reactor vessel inspection systems: development of data acquisition and analysis system for the nuclear vessel weld

Energy Technology Data Exchange (ETDEWEB)

Park, C. H.; Lim, H. T.; Um, B. G. [Korea Advanced Institute of Science and Technology, Taejeon (Korea)

2001-03-01

The objective of this project is to develop an automated ultrasonic data acquisition and data analysis system to examine the reactor vessel weldsIn order to examine nuclear vessel welds including reactor pressure vessel(RPV), huge amount of ultrasonic data from 6 channels should be able to be on-line processed. In addition, ultrasonic transducer scanning device should be remotely controlled, because working place is high radiation area. This kind of an automated ultrasonic testing equipment has not been developed domestically yet In order to develop an automated ultrasonic testing system, RPV ultrasonic testing equipments developed in foreign countries were investigated and the capability of high speed ultrasonic signal processing hardwares was analyzed in this study, ultrasonic signal processing system was designed. And also, ultrasonic data acquisition and analysis software was developed. 11 refs., 6 figs., 9 tabs. (Author)

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

International Nuclear Information System (INIS)

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

1988-01-01

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

Fuel assembly for FBR type reactor and reactor core thereof

International Nuclear Information System (INIS)

Kobayashi, Kaoru.

1998-01-01

The present invention provides a fuel assembly to be loaded to a reactor core of a large sized FBR type reactor, in which a coolant density coefficient can be reduced without causing power peaking in the peripheral region of neutron moderators loaded in the reactor core. Namely, the fuel assembly for the FBR type reactor comprises a plurality of fission product-loaded fuel rods and a plurality of fertile material-loaded fuel rods and one or more rods loading neutron moderators. In this case, the plurality of fertile material-loaded fuel rods are disposed to the peripheral region of the neutron moderator-loaded rods. The plurality of fission product-loaded fuel rods are disposed surrounding the peripheral region of the plurality of fertile material-loaded fuel rods. The neutron moderator comprises zirconium hydride, yttrium hydride and calcium hydride. The fission products are mixed oxide fuels. The fertile material comprises depleted uranium or natural uranium. (I.S.)

Reactor pressure vessel embrittlement of NPP borssele: Design lifetime and lifetime extension

International Nuclear Information System (INIS)

Blom, F.J.

2007-01-01

Embrittlement of the reactor pressure vessel of the Borssele nuclear power plant has been investigated taking account of the design lifetime of 40 years and considering 20 years subsequent lifetime extension. The paper presents the current licensing status based on considerations of material test data and of US nuclear regulatory standards. Embrittlement status is also evaluated against German and French nuclear safety standards. Results from previous fracture toughness and Charpy tests are investigated by means of the Master curve toughness transition approach. Finally, state of the art insights are investigated by means of literature research. Regarding the embrittlement status of the reactor pressure vessel of Borssele nuclear power plant it is concluded that there is a profound basis for the current license up to the original end of the design life in 2013. The embrittlement temperature changes only slightly with respect to the acceptance criterion adopted postulating further operation up to 2033. Continued safe operation and further lifetime extension are therefore not restricted by reactor pressure vessel embrittlement

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

International Nuclear Information System (INIS)

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

1990-01-01

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

Assessment of the TRINO reactor pressure vessel integrity: theoretical analysis and NDE

Energy Technology Data Exchange (ETDEWEB)

Milella, P P; Pini, A [ENEA, Rome (Italy)

1988-12-31

This document presents the method used for the capability assessment of the Trino reactor pressure vessel. The vessel integrity assessment is divided into the following parts: transients evaluation and selection, fluence estimate for the projected end of life of the vessel, characterization of unirradiated and irradiated materials, thermal and stress analysis, fracture mechanics analysis and eventually fracture input to Non Destructive Examination (NDE). For each part, results are provided. (TEC).

The 5th surveillance testing for Kori unit 2 reactor vessel materials

Energy Technology Data Exchange (ETDEWEB)

Chang, Kee Ok; Kim, Byoung Chul; Lee, Sam Lai; Choi, Kwon Jae; Gong, Un Sik; Chang, Jong Hwa; Joo, Yong Sun; Ahn, Sang Bok; Hong, Joon Hwa [Korea Atomic Energy Research Institute, Taejeon (Korea)

2001-03-01

Surveillance testing for reactor vessel materials is performed in order to evaluate the irradiation embrittlement due to neutrons during operation and set the condition of safe operation of nuclear reactor. The 5th surveillance testing was performed completely by Korea Atomic Energy Research Institute at Taejeon after the capsule was transported from Kori site including its removal from reactor. Fast neutron fluences for capsules were calculated and various testing including mechanical and chemistry analysis were performed in order to evaluate the integrity of Kori unit 2 reactor vessel during the operation until life time. The evaluation results are as follows; Fast neutron fluences for capsules V, R, P, T and N are 2.837E+18, 1.105E+19, 2.110E+19, 3.705E+19 and 4.831E+19n/cm{sup 2}, respectively. The bias factor, the ratio of measurement/calculation, was 0.918 for the 1st through 5th testing and the calculational uncertainty, 11.6% satisfied the requirement of USNRC Reg.Guide DG-1053, 20%. The best estimated neutron fluence for reactor vessel inside surface was 1.898E+19n/cm{sup 2} based on the end of 15th fuel cycle and it was predicted that the fluences of vessel inside surface at 32, 40, 48 and 56EFPY would reach 4.203E+19, 5.232E+19, 6.262E+19 and 7.291E+19n/cm{sup 2} based on the current calculation. The result through this analysis for Kori unit 2 showed that there would be no problem for the pressurized thermal shock(PTS) during the operation until design life. 49 refs., 35 figs., 48 tabs. (Author)

Preventive protection device and method for bottom of reactor pressure vessel

International Nuclear Information System (INIS)

Hayashi, Eisaku; Kurosawa, Koichi; Furukawa, Hideyasu; Morinaka, Ren; Enomoto, Kunio; Otaka, Masahiro; Yoshikubo, Fujio; Chiba, Noboru; Sato, Kazunori.

1995-01-01

In a preventive protection device for improving stresses in reactor structural components by jetting highly pressurized water with cavitation bubbles from a jetting nozzle toward structural components in a reactor pressure vessel, a fixed structure to a CRD housing is provided with a rotational body attached to the structure, a multi joint arm and a jetting nozzle supported to the multi joint arm. The jetting nozzle is disposed at a position where the center of the jetting deviates from the center of the CRD housing. In addition, a monitoring camera is disposed for displaying the target for preventive protection. The state of stresses on a plurality of targets for preventive protection can be improved by the preventive protection device at a fixed position in the bottom of a reactor pressure vessel where housings stand densely, thereby enabling to attain the preventive protection operation easily and rapidly. (N.H.)

The coolability limits of a reactor pressure vessel lower head

Energy Technology Data Exchange (ETDEWEB)

Theofanous, T.G.; Syri, S. [Univ. of California, Santa Barbara, CA (United States)

1995-09-01

Configuration II of the ULPU experimental facility is described, and from a comprehensive set of experiments are provided. The facility affords full-scale simulations of the boiling crisis phenomenon on the hemispherical lower head of a reactor pressure vessel submerged in water, and heated internally. Whereas Configuration I experiments (published previously) established the lower limits of coolability under low submergence, pool-boiling conditions, with Configuration II we investigate coolability under conditions more appropriate to practical interest in severe accident management; that is, heat flux shapes (as functions of angular position) representative of a core melt contained by the lower head, full submergence of the reactor pressure vessel, and natural circulation. Critical heat fluxes as a function of the angular position on the lower head are reported and related the observed two-phase flow regimes.

Monitoring PWR reactor vessel liquid level with SPNDs during LOCAs

International Nuclear Information System (INIS)

Adams, J.P.

1982-01-01

Data from in-core self-powered neutron detectors taken during two nuclear loss-of-coolant accident simulations have been correlated with core moderator density changes. The detector current attenuation has been calculated during blowdown and reflood phases of the simulation. Based on these data, it is concluded that these detectors could be used to monitor reactor vessel liquid level during loss-of-coolant accidents in pressurized water reactors

Latest feedback from a major reactor vessel dismantling project

International Nuclear Information System (INIS)

Boucau, J.; Segerud, P.; Sanchez, M.; Garcia, R.

2015-01-01

Westinghouse performed two large segmentation projects in 2010-2013 and then 2013-2015 at the Jose Cabrera nuclear power plant in Spain. The power plant is located in Almonacid de Zorita, 43 miles east of Madrid, Spain and was in operation between 1968 and 2006. This paper will describe the sequential steps required to prepare, segment, separate, and package the individual component segments using under water mechanical techniques. The paper will also include experiences and lessons learned that Westinghouse has collected from the activities performed during the reactor vessel and vessel internals segmentation projects. (authors)

Gamma dose rate estimation and operation management suggestions for decommissioning the reactor pressure vessel of HTR-PM

Energy Technology Data Exchange (ETDEWEB)

Sheng Fang; Hong Li; Jianzhu Cao; Wenqian Li; Feng Xie; Jiejuan Tong [Institute of Nuclear and New Energy Technology, Tsinghua, University, Beijing (China)

2013-07-01

China is now designing and constructing a high temperature gas cooled reactor-pebble bed module (HTR-PM). In order to investigate the future decommissioning approach and evaluate possible radiation dose, gamma dose rate near the reactor pressure vessel was calculated for different cooling durations using QAD-CGA program. The source term of this calculation was provided by KORIGEN program. Based on the calculated results, the spatial distribution and temporal changes of gamma dose rate near reactor pressure vessel was systematically analyzed. A suggestion on planning decommissioning operation of reactor pressure vessel of HTRPM was given based on calculated dose rate and the Chinese Standard GB18871-2002. (authors)

High H⁻ ionic conductivity in barium hydride.

Science.gov (United States)

Verbraeken, Maarten C; Cheung, Chaksum; Suard, Emmanuelle; Irvine, John T S

2015-01-01

With hydrogen being seen as a key renewable energy vector, the search for materials exhibiting fast hydrogen transport becomes ever more important. Not only do hydrogen storage materials require high mobility of hydrogen in the solid state, but the efficiency of electrochemical devices is also largely determined by fast ionic transport. Although the heavy alkaline-earth hydrides are of limited interest for their hydrogen storage potential, owing to low gravimetric densities, their ionic nature may prove useful in new electrochemical applications, especially as an ionically conducting electrolyte material. Here we show that barium hydride shows fast pure ionic transport of hydride ions (H(-)) in the high-temperature, high-symmetry phase. Although some conductivity studies have been reported on related materials previously, the nature of the charge carriers has not been determined. BaH2 gives rise to hydride ion conductivity of 0.2 S cm(-1) at 630 °C. This is an order of magnitude larger than that of state-of-the-art proton-conducting perovskites or oxide ion conductors at this temperature. These results suggest that the alkaline-earth hydrides form an important new family of materials, with potential use in a number of applications, such as separation membranes, electrochemical reactors and so on.

Numerical effects in the neutron flux calculations into WWER-type reactor vessels by Monte Carlo method

International Nuclear Information System (INIS)

Alvarez Cardona, C.M.; Rodriguez Gual, M.; Hernandez Valle, S.

2001-01-01

The calculation of neutron fluxes and fluence into reactor pressure vessel is a regulatory requirement in the stages of the design, operation and plan lifetime extension. The reactor vessel is considered a unique and non-substitutable part of the NPP that undergoes degradation. The main source of the aging comes from the fast neutron damage induced in the steel crystalline lattice. Due to the proximity of the core edge to the vessel inner surface; the vessel steel is exposed to high fast neutron fluence. The effect of this irradiation on the mechanical properties becomes more acute because of the impurities measured in the Russian steel alloys. In the present paper, a PC version of the Monte Carlo 3-D HEXANN-EVALU system is used for the estimation of the WWER reactor pressure vessel irradiation. It was selected on the basis of its flexible options that on the other hand need to be quantified in connection with the desired magnitudes. The parameters that control the random walk of neutrons as well as the efficiency increasing options included in the code are studied in order to identify their impact in the final results for fluxes and fluence in the reactor pressure vessel. As a result an optimal set of parameters is suggested. (authors)

System for bearing a nuclear reactor vessel cooled by liquid metal

International Nuclear Information System (INIS)

Mahe, A.; Jullien, G.

1976-01-01

The invention relates to a bearing system for supporting a nuclear reactor vessel of the kind which is suspended from the reactor closure slab. The bearing system comprises a ring connected at one end to a collar and at the other end to two collars. The collar connected to the bottom end of the ring forms the top part of the vessel to be supported while the other two collars fit into the slab at two separate places. The ring and collars are disposed in an annular space formed in the slab and dividing it into two parts, i.e., a central part and a peripheral part surrounding the central part of the slab

Evaluation of thermal ratcheting of reactor vessel wall near the sodium surface

International Nuclear Information System (INIS)

Take, Kohji; Fujioka, Terutaka; Yano, Kazutaka

1989-01-01

Plastic ratcheting of reactor vessels may occur by an axially moving thermal gradient without primary stress. So there is a need to establish a proper prediction method for the plastic ratcheting. In this study, inelastic FEM analyses of reactor vessel model by using an advanced constitutive equation were carried out in order to comprehend plastic ratcheting behaviour of cylinder which subject to an axially moving thermal gradient. As a result of analyses, a basic mechanism of this ratcheting was found. And it also indicated that cyclic hardening behaviour will became important for development of evaluation method. (author)

Stud bolt handling equipment for reactor vessel

International Nuclear Information System (INIS)

Bunyan, T.W.

1989-01-01

Reactor vessel stud bolt handling equipment includes means for transferring a stud bolt to a carrier from a parking station, or vice versa. Preferably a number of stud bolts are handled simultaneously. The transfer means may include cross arms rotatable about extendable columns, and the equipment is mounted on a mobile base for movement into and out of position. Each carrier comprises a tubular socket and an expandable sleeve to grip a stud bolt. (author)

Thermo-mechanical behaviour of FBTR reactor vessel due to natural convection in cover gas space

International Nuclear Information System (INIS)

Srinivasan, G.; Varadarajan, S.; Kapoor, R.P.

1988-01-01

Fast Breeder Test Reactor is a 40 MW(t), loop type sodium cooled reactor, similar in design to Rapsodie. The Reactor Assembly, which is the heart of FBTR, comprises the Reactor Vessel (RV) housed in a safety vessel within a concrete cell (A1 Cell). The RV which supports the core is shielded at the top by two rotatable plugs which are stacked with layers of borated graphite and steel. The smaller plug (SRP), is mounted excentric to the larger one (LRP). A nominal annular gap of 16 mm is provided between RV and LRP and between LRP and SRP to enable free rotation of the plugs. Stainless Steel insulation is fixed inside the steel vessel, to avoid overheating of the A1 Cell concrete. The core is supported by the Grid Plate (GP), bolted to the RV. During preheating, sodium charging and isothermal runs upto 350 0 C, temperature asymmetries were noticed in the reactor vessel wall in the cover gas space. This was attributable to convection currents in the annulus between RV and LRP. The asymmetries also resulted in a lateral shift of the grid plate. This paper discusses our experience in suppressing these convection currents, and minimising the grid plate shift

Cast iron as structural material for hot-working reactor vessels (PCIV)

International Nuclear Information System (INIS)

Ostendorf, H.; Schmidt, G.; Pittack, W.

1977-01-01

Cast iron with lamellar graphite is best suited for prestressed structures, because its compressive strength is nearly 4 times its tensile strength. In comparison to room temperature, cast iron with lamellar graphite shows essentially no loss of strength up to temperatures of 400 0 C. Under the particular aspect to use cast iron for hot-working prestressed reactor pressure vessels (PCIV) (Prestressed cast iron vessel=PCIV) a materials testing program is carried out, which meets the strict certification requirements for materials in the construction of reactor pressure vessels and which completes the presently available knowledge of cast iron. Especially in the following fields an extension and supplement of the present level of knowledge is necessary. - Mechanical properties under compressive stresses. - Material properties at elevated temperatures. - Influence of irradiation on mechanical and physical properties. - Production standards and quality control. The state of the research and the available data of the material testing program are reported. (Auth.)

Cast iron as structural material for hot-working reactor vessels (PCIV)

International Nuclear Information System (INIS)

Ostendorf, H.; Schmidt, G.; Pittack, W.

1977-01-01

Cast iron with lamellar graphite is best suited for prestressed structures, because its compressive strength is nearly 4 times its tensile strength. In comparison to room temperature, cast iron with lamellar graphite shows essentially no loss of strength up to temperatures of 400 0 C. Under the particular aspect to use cast iron for hot-working prestressed reactor pressure vessels (PCIV) (Prestressed cast iron vessel=PCIV) a materials testing program is carried out, which meets the strict certification requirements for materials in the construction of reactor pressure vessels and which completes the presently available knowledge of cast iron. Especially in the following fields an extension and supplement of the present level of knowledge is necessary: mechanical properties under compressive stresses; material properties at elevated temperatures; influence of irradiation on mechanical and physical properties; production standards and quality control. The state of the research and the available data of the material testing program are reported

Midland reactor pressure vessel flaw distribution

International Nuclear Information System (INIS)

Foulds, J.R.; Kennedy, E.L.; Rosinski, S.T.

1993-12-01

The results of laboratory nondestructive examination (NDE), and destructive cross-sectioning of selected weldment sections of the Midland reactor pressure vessel were analyzed per a previously developed methodology in order to develop a flaw distribution. The flaw distributions developed from the NDE results obtained by two different ultrasonic test (UT) inspections (Electric Power Research Institute NDE Center and Pacific Northwest Laboratories) were not statistically significantly different. However, the distribution developed from the NDE Center's (destructive) cross-sectioning-based data was found to be significantly different than those obtained through the UT inspections. A fracture mechanics-based comparison of the flaw distributions showed that the cross-sectioning-based data, conservatively interpreted (all defects considered as flaws), gave a significantly lower vessel failure probability when compared with the failure probability values obtained using the UT-based distributions. Given that the cross-sectioning data were reportedly biased toward larger, more significant-appearing (by UT) indications, it is concluded that the nondestructive examinations produced definitively conservative results. In addition to the Midland vessel inspection-related analyses, a set of twenty-seven numerical simulations, designed to provide a preliminary quantitative assessment of the accuracy of the flaw distribution method used here, were conducted. The calculations showed that, in more than half the cases, the analysis produced reasonably accurate predictions

Manufacturing and investigation of U-Mo LEU fuel granules by hydride-dehydride processing

International Nuclear Information System (INIS)

Stetskiy, Y.A.; Trifonov, Y.I.; Mitrofanov, A.V.; Samarin, V.I.

2002-01-01

Investigations of hydride-dehydride processing for comminution of U-Mo alloys with Mo content in the range 1.9/9.2% have been performed. Some regularities of the process as a function of Mo content have been determined as well as some parameters elaborated. Hydride-dehydride processing has been shown to provide necessary phase and chemical compositions of U-Mo fuel granules to be used in disperse fuel elements for research reactors. Pin type disperse mini-fuel elements for irradiation tests in the loop of 'MIR' reactor (Dmitrovgrad) have been fabricated using U-Mo LEU fuel granules obtained by hydride-dehydride processing. Irradiation tests of these mini-fuel elements loaded to 4 g U tot /cm 3 are planned to start by the end of this year. (author)

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

International Nuclear Information System (INIS)

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

1988-01-01

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

ACCEPTABILITY ENVELOPE FOR METAL HYDRIDE-BASED HYDROGEN STORAGE SYSTEMS

Energy Technology Data Exchange (ETDEWEB)

Hardy, B.; Corgnale, C.; Tamburello, D.; Garrison, S.; Anton, D.

2011-07-18

The design and evaluation of media based hydrogen storage systems requires the use of detailed numerical models and experimental studies, with significant amount of time and monetary investment. Thus a scoping tool, referred to as the Acceptability Envelope, was developed to screen preliminary candidate media and storage vessel designs, identifying the range of chemical, physical and geometrical parameters for the coupled media and storage vessel system that allow it to meet performance targets. The model which underpins the analysis allows simplifying the storage system, thus resulting in one input-one output scheme, by grouping of selected quantities. Two cases have been analyzed and results are presented here. In the first application the DOE technical targets (Year 2010, Year 2015 and Ultimate) are used to determine the range of parameters required for the metal hydride media and storage vessel. In the second case the most promising metal hydrides available are compared, highlighting the potential of storage systems, utilizing them, to achieve 40% of the 2010 DOE technical target. Results show that systems based on Li-Mg media have the best potential to attain these performance targets.

Pressure vessel codes: Their application to nuclear reactor systems

International Nuclear Information System (INIS)

1966-01-01

A survey has been made by the International Atomic Energy Agency of how the problems of applying national pressure vessel codes to nuclear reactor systems have been treated in those Member States that had pressurized reactors in operation or under construction at the beginning of 1963. Fifteen answers received to an official inquiry form the basis of this report, which also takes into account some recently published material. Although the answers to the inquiry in some cases data back to 1963 and also reflect the difficulty of describing local situations in answer to standard questions, it is hoped that the report will be of interest to reactor engineers. 21 refs, 1 fig., 2 tabs

Radiation effects on reactor pressure vessel supports

International Nuclear Information System (INIS)

Johnson, R.E.

1996-05-01

The purpose of this report is to present the findings from the work done in accordance with the Task Action Plan developed to resolve the Nuclear Regulatory Commission (NRC) Generic Safety Issue No. 15, (GSI-15). GSI-15 was established to evaluate the potential for low-temperature, low-flux-level neutron irradiation to embrittle reactor pressure vessel (RPV) supports to the point of compromising plant safety. An evaluation of surveillance samples from the High Flux Isotope Reactor (HFIR) at the Oak Ridge National Laboratory (ORNL) had suggested that some materials used for RPV supports in pressurized-water reactors could exhibit higher than expected embrittlement rates. However, further tests designed to evaluate the applicability of the HFIR data to reactor RPV supports under operating conditions led to the conclusion that RPV supports could be evaluated using traditional method. It was found that the unique HFIR radiation environment allowed the gamma radiation to contribute significantly to the embrittlement. The shielding provided by the thick steel RPV shell ensures that degradation of RPV supports from gamma irradiation is improbable or minimal. The findings reported herein were used, in part, as the basis for technical resolution of the issue

Evaluation of a cavity flooding strategy for the prevention of reactor vessel failure in a severe accident

Energy Technology Data Exchange (ETDEWEB)

Park, Rae Joon; Je, Moo Sung; Park, Chang Kyoo [Korea Atomic Energy Research Institute, TaeJon (Korea, Republic of)

1994-10-01

As a part of the evaluation of accident management strategies for severe accident prevention or mitigation in a station blackout scenario for YGN 3 and 4, an external vessel cooling strategy for the prevention of reactor vessel failure has been estimated using the MAAP4 computer code. The sensitivity studies have been performed such as actuating timings and the number of spray pumps used. To explore external vessel cooling strategies, containment spray pumps were actuated by varying time spanning core uncovery, core melting and relocation of molten core material. It was shown that flooding of the reactor cavity using the containment spray system may prevent reactor vessel failure but may not prevent the failure of the relocation of molten core material during the station blackout sequence of YGN 3 and 4. Reactor vessel failure can be prevented by external vessel cooling using condensed water from the operation of two containment spray pumps at the time of core melting and using water from the operation of one containment spray pumps at the time of core melting and using water from the operation of one containment spray pump at the time of core uncovery. (Author) 46 refs., 26 figs., 5 tabs.

Stress analyses for reactor pressure vessels by the example of a product line '69 boiling water reactor

International Nuclear Information System (INIS)

Mkrtchyan, Lilit; Schau, Henry; Wolf, Werner; Holzer, Wieland; Wernicke, Robert; Trieglaff, Ralf

2011-01-01

The reactor pressure vessels (RPV) of boiling water reactors (BWR) belonging to the product line '69 have unusually designed heads. The spherical cap-shaped bottom head of the vessel is welded directly to the support flange of the lower shell course. This unusual construction has led repeatedly to controversial discussions concerning the limits and admissibility of stress intensities arising in the junction of the bottom head to the cylindrical shell. In the present paper, stress analyses for the design conditions are performed with the finite element method in order to determine and categorize the occurring stresses. The procedure of stress classification in accordance with the guidelines of German KTA 3201.2 and Section III of the ASME Code (Subsection NB) is described and subsequently demonstrated by the example of a typical BWR vessel. The accomplished investigations yield allowable stress intensities in the considered area. Additionally, limit load analyses are carried out to verify the obtained results. Complementary studies, performed for a torispherical head, prove that the determined maximum peak stresses in the junction between the bottom head and the cylindrical shell are not unusual also for pressure vessels with regular bottom head constructions. (orig.)

Dual shell pressure balanced reactor vessel. Final project report

International Nuclear Information System (INIS)

Robertus, R.J.; Fassbender, A.G.

1994-10-01

The Department of Energy's Office of Energy Research (OER) has previously provided support for the development of several chemical processes, including supercritical water oxidation, liquefaction, and aqueous hazardous waste destruction, where chemical and phase transformations are conducted at high pressure and temperature. These and many other commercial processes require a pressure vessel capable of operating in a corrosive environment where safety and economy are important requirements. Pacific Northwest Laboratory (PNL) engineers have recently developed and patented (U.S. patent 5,167,930 December 1, 1992) a concept for a novel Dual Shell Pressure Balanced Vessel (DSPBV) which could solve a number of these problems. The technology could be immediately useful in continuing commercialization of an R ampersand D 100 award-winning technology, Sludge-to-oil Reactor System (STORS), originally developed through funding by OER. Innotek Corporation is a small business that would be one logical end-user of the DSPBV reactor technology. Innotek is working with several major U.S. engineering firms to evaluate the potential of this technology in the disposal of wastes from sewage treatment plants. PNL entered into a CRADA with Innotek to build a bench-scale demonstration reactor and test the system to advance the economic feasibility of a variety of high pressure chemical processes. Hydrothermal processing of corrosive substances on a large scale can now be made significantly safer and more economical through use of the DSPBV. Hydrothermal chemical reactions such as wet-air oxidation and supercritical water oxidation occur in a highly corrosive environment inside a pressure vessel. Average corrosion rates from 23 to 80 miles per year have been reported by Rice (1994) and Latanision (1993)

Development and investigation of the prestressed reinforced concrete vessels for the water cooled reactors in the FRG

International Nuclear Information System (INIS)

Medovikov, A.I.; Bogopol'skij, V.G.; Nikolaev, Yu.B.; Konevskij, V.N.

1980-01-01

An analysis of calculation results for characteristics of stress-strained state of reactor vessel made of prestressed reinforced concrete is presented. Experimental data obtained during the investigation into a model of reactor vessel top cover are given. Thermal shielding system both for boiling water and pressurized-water reactors has been considered and its working capacity has been evaluated. An analysis of experimental data show correctness of the method assumed for calculation of the reactor top cover which permits to exactly determine its stressed-strained state as well as the nature of crack propagation in the vessel and the structure supporting power. Ceramics is suggested to be used as a heat-insulating material

Microstructural evolution in reactor pressure vessel steel under neutron irradiation

Energy Technology Data Exchange (ETDEWEB)

Ohno, Katsumi; Fukuya, Koji [Institute of Nuclear Safety System Inc., Seika, Kyoto (Japan)

2000-09-01

Understanding microstructural changes in reactor pressure vessel steels is important in order to evaluate radiation-induced embrittlement, one of the major aging phenomena affecting the extension of plant life. In this study, actual surveillance test specimens and samples of rector vessel low-alloy steel (A533B steel) irradiated in a research reactor were examined using state-of-the-art techniques to clarify the neutron flux effect on the microstructural changes. These techniques included small angle neutron scattering and atom probes. Microstructural changes which are considered to be the main factors affecting embrittlement, including the production of copper-rich precipitates and the segregation of impurity elements, were confirmed by the results of the study. In addition, the mechanical properties were predicted based on the obtained quantitative data such as the diameters of precipitates. Consequently, the hardening due to irradiation was almost simulated. (author)

Analytical evaluation of two-phase natural circulation flow characteristics under external reactor vessel cooling

International Nuclear Information System (INIS)

Park, Jong Woon

2009-01-01

This work proposes an analytical method of evaluating the effects of design and operating parameters on the low-pressure two-phase natural circulation flow through the annular shaped gap at the reactor vessel exterior surface heated by corium (molten core) relocated to the reactor vessel lower plenum after loss of coolant accidents. A natural circulation flow velocity equation derived from steady-state mass, momentum, and energy conservation equations for homogeneous two-phase flow is numerically solved for the core melting conditions of the APR1400 reactor. The solution is compared with existing experiments which measured natural circulation flow through the annular gap slice model. Two kinds of parameters are considered for this analytical method. One is the thermal-hydraulic conditions such as thermal power of corium, pressure and inlet subcooling. The others are those for the thermal insulation system design for the purpose of providing natural circulation flow path outside the reactor vessel: inlet flow area, annular gap clearance and system resistance. A computer program NCIRC is developed for the numerical solution of the implicit flow velocity equation.

Experimental investigation of creep behavior of reactor vessel lower head

International Nuclear Information System (INIS)

Chu, T.Y.; Pilch, M.; Bentz, J.H.; Behbahani, A.

1998-03-01

The objective of the USNRC supported Lower Head Failure (LHF) Experiment Program at Sandia National Laboratories is to experimentally investigate and characterize the failure of the reactor pressure vessel (RPV) lower head due to the thermal and pressure loads of a severe accident. The experimental program is complemented by a modeling program focused on the development of a constitutive formulation for use in standard finite element structure mechanics codes. The problem is of importance because: lower head failure defines the initial conditions of all ex-vessel events; the inability of state-of-the-art models to simulate the result of the TMI-II accident (Stickler, et al. 1993); and TMI-II results suggest the possibility of in-vessel cooling, and creep deformation may be a precursor to water ingression leading to in-vessel cooling

Economic analysis of hydride fueled BWR

International Nuclear Information System (INIS)

Ganda, F.; Shuffler, C.; Greenspan, E.; Todreas, N.

2009-01-01

The economic implications of designing BWR cores with hydride fuels instead of conventional oxide fuels are analyzed. The economic analysis methodology adopted is based on the lifetime levelized cost of electricity (COE). Bracketing values (1970 and 3010 $/kWe) are used for the overnight construction costs and for the power scaling factors (0.4 and 0.8) that correlate between a change in the capital cost to a change in the power level. It is concluded that a newly constructed BWR reactor could substantially benefit from the use of 10 x 10 hydride fuel bundles instead of 10 x 10 oxide fuel bundles design presently in use. The cost saving would depend on the core pressure drop constraint that can be implemented in newly constructed BWRs - it is between 2% and 3% for a core pressure drop constraint as of the reference BWR, between 9% and 15% for a 50% higher core pressure drop, and between 12% and 21% higher for close to 100% core pressure. The attainable cost reduction was found insensitive to the specific construction cost but strongly dependent on the power scaling factor. The cost advantage of hydride fuelled cores as compared to that of the oxide reference core depends only weakly on the uranium and SWU prices, on the 'per volume base' fabrication cost of hydride fuels, and on the discount rate used. To be economically competitive, the uranium enrichment required for the hydride fuelled core needs to be around 10%.

Marine reactor pressure vessels dumped in the Kara Sea

International Nuclear Information System (INIS)

Mount, M.E.

1997-01-01

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

Safety of light-water reactor pressure vessels against brittle fracture

International Nuclear Information System (INIS)

Brumovsky, M.

1979-01-01

The results are surveyed of research by SKODA Trust into brittle failure resistance of materials for WWER type reactor pressure vessels and into pressure vessel operating safety. Conditions are discussed in detail decisive for initiation, propagation and arrest of brittle fracture. The tests on the Cr-Mo-V type steel showed high resistance of the steel to the formation and the propagation of brittle fracture. They also confirmed the high operating reliability and the required service life of the steel. (B.S.)

Power reactor embrittlement data base (PR-EDB): Uses in evaluating radiation embrittlement of reactor vessels

International Nuclear Information System (INIS)

Kam, F.B.K.; Stallmann, F.W.; Wang, J.A.

1992-01-01

Investigations of regulatory issues such as vessel integrity over plant life, vessel failure, and sufficiency of current Codes, Standard Review Plans (SRPs), and Guides for license renewal can be greatly expedited by the use of a well-designed, computerized data base. Also, such a data is essential for the evaluation of embrittlement prediction models by researchers. The Power Reactor Embrittlement Data Base (PR-EDB) is such a comprehensive collection of data for US commercial nuclear reactors. The current compilation contains data from 92 reactors and consists of 175 data points for weld materials (79 different welds) and 395 data points for base materials (110 different base materials). The different types of data that are implemented or planned for this data base are discussed. ''User-friendly'' utility programs have been written to investigate a list of problems using this data base. The utility programs are also used to add and upgrade data, retrieve and select specific data, manipulate data, display data to the screen or printer, and to fit and plot Charpy impact data. The results of several studies investigated are presented in this paper

Finite element analysis of prestressed concrete reactor vessels

International Nuclear Information System (INIS)

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

1977-01-01

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

Radiation field analyses in reactor vessels of PWRs

Energy Technology Data Exchange (ETDEWEB)

Fukuya, Koji; Nakata, Hayato; Fujii, Katsuhiko; Kimura, Itsuro [Institute of Nuclear Safety System, Inc., Mihama, Fukui (Japan); Ohmura, Masaki; Kitagawa, Hideo [Mitsubishi Heavy Industries, Ltd., Nuclear Energy Systems Engineering Center, Yokohama, Kanagawa (Japan); Itoh, Taku; Shin, Kazuo [Kyoto Univ. (Japan). Faculty of Engineering

2002-09-01

Radiation analysis in reactor vessels of PWRs were performed using three calculation codes (two dimensional transport code DORT, three dimensional transport code TORT and three dimensional Monte Carlo code MCNP) and three cross section data (ENDF/B-IV, ENDF/B-VI and JENDL3.2) to improve accuracy of estimation for neutron flux, gamma-ray flux and displacement per atom (dpa). The calculations using DORT at a surveillance position agreed with the dosimetry measurements for the three cross sections. The calculated neutron spectra using the three cross sections at the reactor vessels and the surveillance position were quite similar to each other. The difference in the cross sections gave small impacts on the fluence estimation. The ratio of the calculations to the measurements using TORT was similar to those using DORT, indicating that TORT is applicable to the radiation analysis in PWRs. The MCNP calculations resulted in a similar agreement with the dosimeter measurement to the DORT calculation while they needed a long computing time. Improvement of calculation techniques is needed for application of MCNP. The calculated dpa agreed within 10% for the three cross sections. (author)

An assessment of the economic consequences of thermal annealing of a nuclear reactor pressure vessel

International Nuclear Information System (INIS)

Griesbach, T.J.; Server, W.L.

1991-01-01

The use of a thermal heat treatment to recover mechanical properties which were degraded by neutron radiation exposure is a potential method for assuring reactor pressure vessel licensing life and possible license renewal. 'Wet anneals' at temperatures less than 343degC have been conducted on test reactors in Alaska (SM-1A) and Belgium (BR3). The Soviets have also performed 'dry anneals' at higher temperatures near or above 450degC on several commercial reactor vessels. Technical and economic uncertainties have made utilities in the United States reluctant to seriously consider thermal annealing of large commercial reactor vessels except as a last resort option. However, as a utility begins to experience significant radiation embrittlement or considers extending the operating license life of the vessel, thermal annealing can be a viable option depending upon many considerations. These considerations include other possible remedial measures that can be taken (i.e., flux reduction), economic issues with regard to utility finances, and corporate philosophy. A decision analysis model has been developed to analyze the thermal anneal option in comparison to other alternatives for a number of possible combinations and timing. The results for a postulated vessel and embrittlement condition are presented to show that thermal annealing can be a viable management option which should be taken seriously. (author)

The development of reactor vessel internal heavy forging for 1000 MW pressurized-water reactor nuclear power plant

International Nuclear Information System (INIS)

Zhang Zhifeng; Chen Yongbo; Ding Xiuping; Zhang Lingfang

2012-01-01

This Paper introduced the development of Reactor Vessel Internal (RVI) heavy forgings for 1000 MW Pressurized Water Reactor (PWR) nuclear power plant, analyzed the manufacture difficulties and technical countermeasures. The testing result of the product indicated that the performance of RVI heavy forgings manufactured by Shanghai Heavy Machinery Plant Ld. (SHMP) is outstanding and entirely satisfy the technical requirements for RVI product. (authors)

Proving Test on the Reliability for Reactor Containment Vessel

International Nuclear Information System (INIS)

Takumi, K.; Nonaka, A.

1988-01-01

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

Consequence evaluation of radiation embrittlement of Trojan reactor pressure vessel supports

International Nuclear Information System (INIS)

Lu, S.C.; Sommer, S.C.; Johnson, G.L.; Lambert, H.E.

1990-10-01

This report describes a consequence evaluation to address safety concerns raised by the radiation embrittlement of the reactor pressure vessel (RPV) supports for the Trojan nuclear power plant. The study comprises a structural evaluation and an effects evaluation and assumes that all four reactor vessel supports have completely lost the load carrying capability. By demonstrating that the ASME code requirements governing Level D service limits are satisfied, the structural evaluation concludes that the Trojan reactor coolant loop (RCL) piping is capable of transferring loads to the steam generator (SG) supports and the reactor coolant pump (RCP) supports. A subsequent design margins to accommodate additional loads transferred to them through the RCL piping. The effects evaluation, employing a systems analysis approach, investigates initiating events and the reliability of the engineered safeguard systems as the RPV is subject to movements caused by the RPV support failure. The evaluation identifies a number of areas of additional safety concerns, but further investigation of the above safety concerns, however, concludes that a hypothetical failure of the Trojan RPV supports due to radiation embrittlement will not result in consequences of significant safety concerns

Probabilistic approach to the analysis of reactor pressure vessel integrity during a pressurized thermal shock

International Nuclear Information System (INIS)

Adamec, P.

2000-12-01

Following a general summary of the issue, an overview of international experience (USA; Belgium, France, Germany, Russia, Spain, Sweden, The Netherlands, and the UK; and probabilistic PTS assessment for the reactor pressure vessel at Loviisa-1, Finland) is presented, and the applicable computer codes (VISA-II, OCA-P, FAVOR, ZERBERUS) are highlighted and their applicability to VVER type reactor pressure vessels is outlined. (P.A.)

Evaluation of creep-fatigue crack growth for large-scale FBR reactor vessel and NDE assessment

Energy Technology Data Exchange (ETDEWEB)

Joo, Young Sang; Kim, Jong Bum; Kim, Seok Hun; Yoo, Bong

2001-03-01

Creep fatigue crack growth contributes to the failure of FRB reactor vessels in high temperature condition. In the design stage of reactor vessel, crack growth evaluation is very important to ensure the structural safety and setup the in-service inspection strategy. In this study, creep-fatigue crack growth evaluation has been performed for the semi-elliptical surface cracks subjected to thermal loading. The thermal stress analysis of a large-scale FBR reactor vessel has been carried out for the load conditions. The distributions of axial, radial, hoop, and Von Mises stresses were obtained for the loading conditions. At the maximum point of the axial and hoop stress, the longitudinal and circumferential surface cracks (i.e. PTS crack, NDE short crack and shallow long crack) were postulated. Using the maximum and minimum values of stresses, the creep-fatigue crack growth of the proposed cracks was simulated. The crack growth rate of circumferential cracks becomes greater than that of longitudinal cracks. The total crack growth of the largest PTS crack is very small after 427 cycles. The structural integrity of a large-scale reactor can be maintained for the plant life. The crack depth growth of the shallow long crack is faster than that of the NDE short crack. In the ISI of the large-scale FBR reactor vessel, the ultrasonic inspection is beneficial to detect the shallow circumferential cracks.

Probability of fracture and life extension estimate of the high-flux isotope reactor vessel

International Nuclear Information System (INIS)

Chang, S.J.

1998-01-01

The state of the vessel steel embrittlement as a result of neutron irradiation can be measured by its increase in ductile-brittle transition temperature (DBTT) for fracture, often denoted by RT NDT for carbon steel. This transition temperature can be calibrated by the drop-weight test and, sometimes, by the Charpy impact test. The life extension for the high-flux isotope reactor (HFIR) vessel is calculated by using the method of fracture mechanics that is incorporated with the effect of the DBTT change. The failure probability of the HFIR vessel is limited as the life of the vessel by the reactor core melt probability of 10 -4 . The operating safety of the reactor is ensured by periodic hydrostatic pressure test (hydrotest). The hydrotest is performed in order to determine a safe vessel static pressure. The fracture probability as a result of the hydrostatic pressure test is calculated and is used to determine the life of the vessel. Failure to perform hydrotest imposes the limit on the life of the vessel. The conventional method of fracture probability calculations such as that used by the NRC-sponsored PRAISE CODE and the FAVOR CODE developed in this Laboratory are based on the Monte Carlo simulation. Heavy computations are required. An alternative method of fracture probability calculation by direct probability integration is developed in this paper. The present approach offers simple and expedient ways to obtain numerical results without losing any generality. In this paper, numerical results on (1) the probability of vessel fracture, (2) the hydrotest time interval, and (3) the hydrotest pressure as a result of the DBTT increase are obtained

Transient temperature and stress distributions in the pressure vessel's wall of a nuclear reactor

International Nuclear Information System (INIS)

Silva, G.A. da

1979-01-01

In order to calculate the temperature distribution in a reactor vessel wall which is under the effect of gamma radiation originated in the reactor core, a numerical solution is proposed. This problem may arise from a reactor cooling pump failure .The thermal stresses are also calculated. (Author) [pt

Scoping calculations for design and analysis of large reactor vessels for liquid-metal fast breeder reactor (LMFBR) plants

International Nuclear Information System (INIS)

Fiala, C.; Kulak, R.F.; Ma, D.C.; Pan, Y.C.; Seidensticker, R.W.; Wang, C.Y.; Zeuch, W.R.

1982-01-01

Reactor vessels for commercial-sized LMFBR plants are quite large - ranging 40 to 70 ft in diameter and 50 to 70 ft in overall depth. These stainless steel vessels contain liquid sodium at relatively low pressures, but at high temperatures. The resulting thin-walled vessels present the structural designer and analyst with special problems, particularly in providing a balanced design to accommodate seismic loads, design basis accident loads, and thermal loadings. A comprehensive set of scoping calculations - though preliminary in detail and depth of design - provides substantial guidance to the vessel designer for subsequent design iterations. Emphasis is placed on the analysis of the large-diameter top closure of the vessel - the deck structure

Reliability aspects of radiation damage in reactor pressure vessel mterials

International Nuclear Information System (INIS)

Brumovsky, M.

1985-01-01

The service life estimate is a major factor in the evaluation of the operating reliability and safety of a nuclear reactor pressure vessel. The evaluation of the service life of the pressure vessel is based on a comparison of fracture toughness values with stress intensity factors. Notch toughness curves are used for the indirect determination of fracture toughness. The dominant degradation effect is radiation embrittlement. Factors having the greatest effect on the result are the properties of the starting material of the vessel and the impurity content, mainly the Cu and P content. The design life is affected by the evaluation of residual lifetime which is made by periodical nondestructive inspections and using surveillance samples. (M.D.)

RPV-1: A Virtual Test Reactor to simulate irradiation effects in light water reactor pressure vessel steels

International Nuclear Information System (INIS)

Jumel, Stephanie; Van-Duysen, Jean Claude

2005-01-01

Many key components in commercial nuclear reactors are subject to neutron irradiation which modifies their mechanical properties. So far, the prediction of the in-service behavior and the lifetime of these components has required irradiations in so-called 'Experimental Test Reactors'. This predominantly empirical approach can now be supplemented by the development of physically based computer tools to simulate irradiation effects numerically. The devising of such tools, also called Virtual Test Reactors (VTRs), started in the framework of the REVE Project (REactor for Virtual Experiments). This project is a joint effort among Europe, the United States and Japan aimed at building VTRs able to simulate irradiation effects in pressure vessel steels and internal structures of LWRs. The European team has already built a first VTR, called RPV-1, devised for pressure vessel steels. Its inputs and outputs are similar to those of experimental irradiation programs carried out to assess the in-service behavior of reactor pressure vessels. RPV-1 is made of five codes and two databases which are linked up so as to receive, treat and/or convey data. A user friendly Python interface eases the running of the simulations and the visualization of the results. RPV-1 is sensitive to its inputs (neutron spectrum, temperature, ...) and provides results in conformity with experimental ones. The iterative improvement of RPV-1 has been started by the comparison of simulation results with the database of the IVAR experimental program led by the University of California Santa Barbara. These first successes led 40 European organizations to start developing RPV-2, an advanced version of RPV-1, as well as INTERN-1, a VTR devised to simulate irradiation effects in stainless steels, in a large effort (the PERFECT project) supported by the European Commission in the framework of the 6th Framework Program

Safety Research Experiment Facility Project. Conceptual design report. Volume V. Reactor vessel and closure

International Nuclear Information System (INIS)

1975-12-01

The Prestressed Concrete Reactor Vessel (PCRV) will serve as the primary pressure retaining structure for the Safety Research Experiment Facility (SAREF) reactor. The reactor core, control rod drive room, primary heat exchangers, and gas circulators will be located in cavities within the PCRV. The orientation of these cavities, except for the control rod drive room, will be similar to the high-temperature gas-cooled reactor (HTGR) designs that are currently proposed or under design. Due to the nature of this type of structure, all biological and radiological shielding requirements are incorporated into the basic vessel design. At the midcore plane there are three radially oriented slots that will extend from the outside surface of the PCRV to the reactor core liner. These slots will accommodate each of the fuel motion monitoring systems which will be part of the observation apparatus used with the loop experiments

Nuclear reactor installation with outer shell enclosing a primary pressure vessel

International Nuclear Information System (INIS)

1975-01-01

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

A prestressed concrete pressure vessel for helium high temperature reactor system

International Nuclear Information System (INIS)

Horner, R.M.W.; Hodzic, A.

1976-01-01

A novel prestressed concrete pressure vessel has been developed to provide the primary containment for a fully integrated system comprising a high temperature nuclear reactor, three horizontally mounted helium turbines, associated heat exchangers and inter-connecting ducts. The design and analysis of the pressure vessel is described. Factors affecting the final choice of layout are discussed, and earlier development work seeking to resolve the conflicting requirements of the structural, mechanical, and system engineers outlined. Proposals to increase the present output of about 1000 MW of electrical power to over 3000 MW, by incorporating four turbines in a single pressure vessel are presented. (author)

Allowable minimum upper shelf toughness for nuclear reactor pressure vessels

Energy Technology Data Exchange (ETDEWEB)

Zahoor, A.

1988-05-01

The paper develops methodology and procedure for determining the allowable minimum upper shelf toughness for continued safe operation of nuclear reactor pressure vessels. Elastic-plastic fracture mechanics analysis method based on the J-integral tearing modulus (J/T) approach is used. Closed from expressions for the applied J and tearing modulus are presented for finite length, part-throughwall axial flaw with aspect ratio of 1/6. Solutions are then presented for Section III, Appendix G flaw. A simple flaw evaluation procedure that can be applied quickly by utility engineers is presented. An attractive feature of the simple procedure is that tearing modulus calculations are not required by the user, and a solution for the slope of the applied J/T line is provided. Results for the allowable minimum upper shelf toughness are presented for a range of reactor pressure vessel thickness and heatup/cooldown rates.

Allowable minimum upper shelf toughness for nuclear reactor pressure vessels

International Nuclear Information System (INIS)

Zahoor, A.

1988-01-01

The paper develops methodology and procedure for determining the allowable minimum upper shelf toughness for continued safe operation of nuclear reactor pressure vessels. Elastic-plastic fracture mechanics analysis method based on the J-integral tearing modulus (J/T) approach is used. Closed from expressions for the applied J and tearing modulus are presented for finite length, part-throughwall axial flaw with aspect ratio of 1/6. Solutions are then presented for Section III, Appendix G flaw. A simple flaw evaluation procedure that can be applied quickly by utility engineers is presented. An attractive feature of the simple procedure is that tearing modulus calculations are not required by the user, and a solution for the slope of the applied J/T line is provided. Results for the allowable minimum upper shelf toughness are presented for a range of reactor pressure vessel thickness and heatup/cooldown rates. (orig.)

Research program plan: reactor vessels. Volume 1

International Nuclear Information System (INIS)

Vagins, M.; Taboada, A.

1985-07-01

The ability of the licensing staff of the NRC to make decisions concerning the present and continuing safety of nuclear reactor pressure vessels under both normal and abnormal operating conditions is dependent upon the existence of verified analysis methods and a solid background of applicable experimental data. It is the role of this program to provide both the analytical methods and the experimental data needed. Specifically, this program develops fracture mechanics analysis methods and design criteria for predicting the stress levels and flaw sizes required for crack initiation, propagation, and arrest in LWR pressure vessels under all known and postulated operations conditions. To do this, not only must the methods be developed but they must be experimentally validated. Further, the materials data necessary for input to these analytical methods must be developed. Thus, in addition to methods development and large scale experimental verification this program also develops data to show that slow-load fracture toughness, rapid-load fracture toughness, and crack arrest toughness obtained from small laboratory specimens are truly representative of the toughness characteristics of the material behavior in pressure vessels in both the unirradiated and the irradiated conditions

Zircaloy-4 hydridation

International Nuclear Information System (INIS)

Vizcaino, Pablo

1997-01-01

The objectives of this work can be summarized as: 1) To reproduce, by heat treatments, matrix microstructures and hydride morphologies similar to those observed in structural components of the CNA-1 and CNE nuclear power plants; 2) To study the evolution of the mechanical properties of the original material with different hydrogen concentrations, such as microhardness, and its capacity to distinguish these materials; 3) To find parameters that allow to estimate the hydrogen content of a material by quantitative metallographic techniques, to be used as complementary in the study of the radioactive materials from reactors

Feedback from Westinghouse experience on segmentation of reactor vessel internals - 59013

International Nuclear Information System (INIS)

Kreitman, Paul J.; Boucau, Joseph; Segerud, Per; Fallstroem, Stefan

2012-01-01

With more than 25 years of experience in the development of reactor vessel internals segmentation and packaging technology, Westinghouse has accumulated significant know-how in the reactor dismantling market. Building on tooling concepts and cutting methodologies developed decades ago for the successful removal of nuclear fuel from the damaged Three Mile Island Unit 2 reactor (TMI-2), Westinghouse has continuously improved its approach to internals segmentation and packaging by incorporating lessons learned and best practices into each successive project. Westinghouse has developed several concepts to dismantle reactor internals based on safe and reliable techniques, including plasma arc cutting (PAC), abrasive water-jet cutting (AWJC), metal disintegration machining (MDM), or mechanical cutting. Westinghouse has applied its technology to all types of reactors covering Pressurized Water Reactors (PWR's), Boiling Water Reactors (BWR's), Gas Cooled Reactors (GCR's) and sodium reactors. The primary challenges of a segmentation and packaging project are to separate the highly activated materials from the less-activated materials and package them into appropriate containers for disposal. Since space is almost always a limiting factor it is therefore important to plan and optimize the available room in the segmentation areas. The choice of the optimum cutting technology is important for a successful project implementation and depends on some specific constraints like disposal costs, project schedule, available areas or safety. Detailed 3-D modeling is the basis for tooling design and provides invaluable support in determining the optimum strategy for component cutting and disposal in waste containers, taking account of the radiological and packaging constraints. Westinghouse has also developed a variety of special handling tools, support fixtures, service bridges, water filtration systems, video-monitoring systems and customized rigging, all of which are required for a

Stress analyses for reactor pressure vessels by the example of a product line '69 boiling water reactor

Energy Technology Data Exchange (ETDEWEB)

Mkrtchyan, Lilit; Schau, Henry [TUEV SUED Energietechnik GmbH, Mannheim (Germany). Abt. Strukturverhalten; Wolf, Werner; Holzer, Wieland [TUEV SUED Industrie Service GmbH, Muenchen (Germany). Abt. Behaelter und Turbosatz; Wernicke, Robert; Trieglaff, Ralf [TUEV NORD SysTec GmbH und Co. KG, Hamburg (Germany). Abt. Festigkeit und Konstruktion

2011-08-15

The reactor pressure vessels (RPV) of boiling water reactors (BWR) belonging to the product line '69 have unusually designed heads. The spherical cap-shaped bottom head of the vessel is welded directly to the support flange of the lower shell course. This unusual construction has led repeatedly to controversial discussions concerning the limits and admissibility of stress intensities arising in the junction of the bottom head to the cylindrical shell. In the present paper, stress analyses for the design conditions are performed with the finite element method in order to determine and categorize the occurring stresses. The procedure of stress classification in accordance with the guidelines of German KTA 3201.2 and Section III of the ASME Code (Subsection NB) is described and subsequently demonstrated by the example of a typical BWR vessel. The accomplished investigations yield allowable stress intensities in the considered area. Additionally, limit load analyses are carried out to verify the obtained results. Complementary studies, performed for a torispherical head, prove that the determined maximum peak stresses in the junction between the bottom head and the cylindrical shell are not unusual also for pressure vessels with regular bottom head constructions. (orig.)

Designs for remote inspection of the ALMR Reactor Vessel Auxiliary Cooling System (RVACS)

International Nuclear Information System (INIS)

Sweeney, F.J.; Carroll, D.G.; Chen, C.; Crane, C.; Dalton, R.; Taylor, J.R.; Tosunoglu, S.; Weymouth, T.

1993-01-01

One of the most important safety systems in General Electric's (GI) Advanced Liquid Metal Reactor (ALMR) is the Reactor Vessel Auxiliary Cooling System (RVACS). Because of high temperature, radiation, and restricted space conditions, GI desired methods to remotely inspect the RVACS, emissive coatings, and reactor vessel welds during normal refueling operations. The DOE/NE Robotics for Advanced Reactors program formed a team to evaluate the ALMR design for remote inspection of the RVACS. Conceptual designs for robots to perform the required inspection tasks were developed by the team. Design criteria for these remote systems included robot deployment, power supply, navigation, environmental hardening of components, tether management, communication with an operator, sensing, and failure recovery. The operation of the remote inspection concepts were tested using 3-D simulation models of the ALMR. In addition, the team performed an extensive technology review of robot components that could survive the environmental conditions in the RVACS

BLENDED CALCIUM ALUMINATE-CALCIUM SULFATE CEMENT-BASED GROUT FOR P-REACTOR VESSEL IN-SITU DECOMMISSIONING

Energy Technology Data Exchange (ETDEWEB)

Langton, C.; Stefanko, D.

2011-03-10

The objective of this report is to document laboratory testing of blended calcium aluminate - calcium hemihydrate grouts for P-Reactor vessel in-situ decommissioning. Blended calcium aluminate - calcium hemihydrate cement-based grout was identified as candidate material for filling (physically stabilizing) the 105-P Reactor vessel (RV) because it is less alkaline than portland cement-based grout which has a pH greater than 12.4. In addition, blended calcium aluminate - calcium hemihydrate cement compositions can be formulated such that the primary cementitious phase is a stable crystalline material. A less alkaline material (pH {<=} 10.5) was desired to address a potential materials compatibility issue caused by corrosion of aluminum metal in highly alkaline environments such as that encountered in portland cement grouts [Wiersma, 2009a and b, Wiersma, 2010, and Serrato and Langton, 2010]. Information concerning access points into the P-Reactor vessel and amount of aluminum metal in the vessel is provided elsewhere [Griffin, 2010, Stefanko, 2009 and Wiersma, 2009 and 2010, Bobbitt, 2010, respectively]. Radiolysis calculations are also provided in a separate document [Reyes-Jimenez, 2010].

Review of analysis methods for prestressed concrete reactor vessels

International Nuclear Information System (INIS)

Dodge, W.G.; Bazant, Z.P.; Gallagher, R.H.

1977-02-01

Theoretical and practical aspects of analytical models and numerical procedures for detailed analysis of prestressed concrete reactor vessels are reviewed. Constitutive models and numerical algorithms for time-dependent and nonlinear response of concrete and various methods for modeling crack propagation are discussed. Published comparisons between experimental and theoretical results are used to assess the accuracy of these analytical methods

Reactor vessel closure head replacements in 1997

International Nuclear Information System (INIS)

Anon.

1997-01-01

The Framatome-Jeumont Industrie consortium have completed in 1997 28 reactor vessel (RV) closure head replacements, including five on 1300 MWe class PWR units. Framatome manages the operations and handles removal and reinstallation of equipment (not including the control rod drive mechanisms (CRDM)) and the requalification tests, while JI, which manufactures the CRDMs, is involved in the CRDM cutting, re-machining and welding operations, using tools of original design, in order to optimize the RV closure head operation in terms of costs, schedule and dosage

Data on test results of vessel cooling system of high temperature engineering test reactor

International Nuclear Information System (INIS)

Saikusa, Akio; Nakagawa, Shigeaki; Fujimoto, Nozomu; Tachibana, Yukio; Iyoku, Tatsuo

2003-02-01

High Temperature Engineering Test Reactor (HTTR) is the first graphite-moderated helium gas cooled reactor in Japan. The rise-to-power test of the HTTR started on September 28, 1999 and thermal power of the HTTR reached its full power of 30 MW on December 7, 2001. Vessel Cooling System (VCS) of the HTTR is the first Reactor Cavity Cooling System (RCCS) applied for High Temperature Gas Cooled Reactors. The VCS cools the core indirectly through the reactor pressure vessel to keep core integrity during the loss of core flow accidents such as depressurization accident. Minimum heat removal of the VCS to satisfy its safety requirement is 0.3MW at 30 MW power operation. Through the performance test of the VCS in the rise-to-power test of the HTTR, it was confirmed that the VCS heat removal at 30 MW power operation was higher than 0.3 MW. This paper shows outline of the VCS and test results on the VCS performance. (author)

Concept of a nuclear powered submersible research vessel and a compact reactor

International Nuclear Information System (INIS)

Kusunoki, Tsuyoshi; Odano, Naoteru; Yoritsune, Tsutomu; Ishida, Toshihisa; Nishimura, Hajime; Tokunaga, Sango

2001-07-01

A conceptual design study of a submersible research vessel navigating in 600 m depth and a compact nuclear reactor were carried out for the expansion of the nuclear power utilization. The mission of the vessel is the research of mechanism of the climate change to predict the global environment. Through conditions of the Arctic Ocean and the sea at high latitude have significant impacts on the global environmental change, it is difficult to investigate those areas by ordinary ships because of thick ice or storm. Therefore the research vessel is mainly utilized in the Arctic Ocean and the sea at high latitude. By taking account of the research mission, the basic specifications of the vessel are decided; the total weight is 500 t, the submersible depth is 600 m, the maximum speed is 12 knots (22.2 km/h), and the number of crews is 16. Nuclear power has an advantage in supplying large power of electricity in the sea for long period. Based on the requirements, it has been decided that two sets of submersible compact reactor, SCR, which is light-weighted and of enhanced safety characteristics of supply the total electricity of 500 kW. (author)

Electrolytic hydriding and hydride distribution in zircaloy-4

International Nuclear Information System (INIS)

Gomes, M.H.L.

1974-01-01

A study has been made of the electrolytic hydriding of zircaloy-4 in the range 20-80 0 C, for reaction times from 5 to 30 hours, and the effect of potential, pH and dissolved oxygen has been investigated. The hydriding reaction was more sensitive to time and temperature conditions than to the electrochemical variables. It has been shown that a controlled introduction of hydrides in zircaloy is feasible. Hydrides were found to be plate like shaped and distributed mainly along grain-boundaries. It has been shown that hydriding kinetics do not follow a simple law but may be described by a Johnson-Mehl empirical equation. On the basis of this equation an activation energy of 9.400 cal/mol has been determined, which is close to the activation energy for diffusion of hydrogen in the hydride. (author)

Vessel core seismic interaction for a fast reactor