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Sample records for bwr reactor vessel

  1. 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.)

  2. 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)

  3. 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)

  4. 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)

  5. Study on reactor vessel replacement (RVR) for 1100 MW class BWR plants in Japan

    International Nuclear Information System (INIS)

    Mizutani, J.; Kawamura, S.; Aoki, M.; Mori, T.

    2001-01-01

    Plant Life Management (PLM) is being studied in Japan, and reactor vessel replacement (RVR) is being considered as one option. Since reactor internals, except for reusable parts, and the reactor pressure vessel (RPV) are replaced, the RVR provides an effective technology for extending the service life of nuclear power plants substantially. At ICONE 7, we reported on the technical viability of the RVR for BWR4-type 800 MWe class plants. This time, we rationalized the RVR method through a study for BWR5-type 1100 MWe class plants to reduce the RVR duration and evaluated the technical viability and the economic efficiency of the method. In addition, we discuss how to dispose of the RPV to complete a scenario of the process from the RVR to its final disposal. (author)

  6. Analysis of the integrity of the pressure vessel of the BWR type nuclear reactor

    International Nuclear Information System (INIS)

    Silva Luna, O.

    1982-01-01

    The presssure vessel of a BWR type reactor was monitored for cracking during alternating events of its in-service life. The monitoring was to determine criticality of fractures catastrophic fractures and the velocity of fracture propagation. Detected cracks were evaluated as specified in ASME code section XI, of a minimum wall thickness of 2.5% crack growths were compared a) of 1/10 of the critical maximum size and b) at in-service inspection intervals according to ASME recommendations to be established at the Laguna Verde nuclear plant. Finally conclusions are made and discussed. (author)

  7. 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)

  8. Ultrasonic phased array examination of circumferential weld joint in reactor pressure vessel of BWR

    Energy Technology Data Exchange (ETDEWEB)

    Nanekar, Paritosh, E-mail: pnanekar@barc.gov.in [Quality Assurance Division, Bhabha Atomic Research Centre, Trombay, Mumbai 400085 (India); Jothilakshmi, N. [Quality Assurance Division, Bhabha Atomic Research Centre, Trombay, Mumbai 400085 (India); Jayakumar, T. [Metallurgy and Materials Group, Indira Gandhi Centre for Atomic Research, Kalpakkam (India)

    2013-12-15

    Highlights: • Phased array technique developed for weld joint inspection in BWR pressure vessel. • Simulation studies were carried out for conventional and phased array probe. • Conventional ultrasonic test shows in-adequate weld coverage and poor resolution. • Focused sound beam in phased array results in good resolution and sensitivity. • Ultrasonic phased array technique is validated on mock-up with reference defects. - Abstract: The weld joints in the reactor pressure vessel (RPV) of Boiling Water Reactors (BWR) are required to be examined periodically for assurance of structural integrity. Ultrasonic phased array examination technique has been developed in authors’ laboratory for inspection of the top flange to shell circumferential weld joint in RPV of BWRs, which are in operation in India since the late 1960s. The development involved detailed simulation studies for computation of focal laws followed by validation on mock-up. The paper brings out the limitations of the conventional ultrasonic technique and how this can be overcome by the phased array approach for the weld joint under consideration. The phased array technique was successfully employed for field examination of this weld joint in RPV during the re-fuelling outage.

  9. BWR type reactor

    International Nuclear Information System (INIS)

    Okano, Shigeru.

    1992-01-01

    In a BWR type reactor, control rod drives are disposed in the upper portion of a reactor pressure vessel, and a control rod guide tube is disposed in adjacent with a gas/liquid separator at a same height, as well as a steam separator is disposed in the control rod guide tube. The length of a connection rod can be shortened by so much as the control rod guide tube and the gas/liquid separator overlapping with each other. Since the control rod guide tube and the gas/liquid separator are at the same height, the number of the gas/liquid separators to be disposed is decreased and, accordingly, even if the steam separation performance by the gas/liquid separator is lowered, it can be compensated by the steam separator of the control rod guide tube. In view of the above, since the direction of emergent insertion of the control rod is not against gravitational force but it is downward direction utilizing the gravitational force, reliability for the emergent insertion of the control rod can be further improved. Further, the length of the connection rod can be minimized, thereby enabling to lower the height of the reactor pressure vessel. The construction cost for the nuclear power plant can be reduced. (N.H.)

  10. BWR vessel and internals project (BWRVIP)

    International Nuclear Information System (INIS)

    Bilanin, W.J.; Dyle, R.L.

    1996-01-01

    Recent Boiling Water Reactor (BWR) inspections indicate that Intergranular Stress Corrosion Cracking (IGSCC) is a significant technical issue for some BWR internals. IN response, the Boiling Water Reactor Vessel and Internals Project (BWRVIP) was formed by an associated of domestic and international utilities which own and operate BWRs. The project is identifying or developing generic, cost-effective strategies for managing degradation of reactor internals from which each utility can select the alternative most appropriate for their plant. The Electric Power Research Institute manages the technical program, implementing the utility defined programs. The BWRVIP is organized into four technical tasks: Assessment, Inspection, Repair and Mitigation. An Integration task coordinates the work. The goal of the Assessment task is to develop methodologies for evaluation of vessel and internal components in support of decisions for operation, inspection, mitigation or repair. The goal of the Inspection task is to develop and assess effective and predictable inspection techniques which can be used to determine the condition of BWR vessel and internals that are potentially susceptible to service-related SCC degradation. The goal of the Repair task is to assure the availability of cost-effective repair/replacement alternatives. The goal of the Mitigation task is to develop and demonstrate countermeasures for SCC degradation. This paper summarizes the BWRVIP approach for addressing BWR internals SCC degradation and illustrates how utilities are utilizing BWRVIP products to successfully manage the effect of SCC on core shrouds

  11. BWR type nuclear reactors

    International Nuclear Information System (INIS)

    Yamamoto, Toru.

    1987-01-01

    Purpose: To obtain reactor core characteristics with less changes in the excess reactivity due to fuel burnup even when the operation period varies. Constitution: In a BWR type reactor where fuel assemblies containing fuel rods incorporated with burnable poisons are arranged, the fuel assemblies are grouped into first fuel assemblies and second fuel assemblies. Then, the number of fuel rods incorporated with burnable poisons within the first fuel assemblies is made greater than that of the second fuel rods, while the concentration of the burnable poisons in the fuel rods incorporated with the burnable poisons in the first fuel assemblies is made lower than that of the fuel rods incorporated with the burnable poisons in the second fuel assemblies. In the BWR type reactor constituted in this way, the reactor core characteristics can be improved by changing the ratio between the first fuel assemblies and the second fuel assemblies charged to the reactor core, thereby decreasing the changes in the burnup of the excess reactivity. (Kamimura, M.)

  12. 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

  13. Neutron flux calculation and fluence in the encircling of the core and vessel of a reactor BWR

    International Nuclear Information System (INIS)

    Martinez C, E.

    2011-01-01

    One of the main objectives related to the safety of any nuclear power plant, including the nuclear power plant of Laguna Verde is to ensure the structural integrity of reactor pressure vessel. To identify and quantify the damage caused by neutron irradiation in the vessel of any nuclear reactor, it is necessary to know both the neutron flux and the neutron fluence that the vessel has been receiving during its operation lifetime, and that the damage observed by mechanical testing are products of microstructural effects induced by neutron irradiation; therefore, it is important the study and prediction of the neutron flux in order to have a better understanding of the damage that these materials are receiving. The calculation here described uses the DORT code, which solves the neutron transport equation in discrete ordinates in two dimensions (x-y, r-θ and r-z), according to a regulatory guide, it should make an approximation of the neutron flux in three dimensions by the so called synthesis method. It is called in that way because it achieves a representation of 3 Dimensional neutron flux combining or summarizing the fluxes calculated by DORT r-θ, r-z and r. This work presents the application of synthesis method, according to Regulatory Guide 1190, to determine the 3 Dimensional fluxes in internal BWR reactor using three different spatial meshes. The results of the neutron flux and fluence, using three different meshes in the directions r, θ and z were compared with results reported in the literature obtaining a difference not larger than 9.61%, neutron flux reached its maximum, 1.58 E + 12 n/cm 2 s, at a height H 4 (239.07 cm) and angle 32.236 o in the core shroud and 4.00 E + 09 n/cm 2 s at a height H 4 and angle 35.27 o in the inner wall of the reactor vessel, positions that are consistent to within ±10% over the ones reported in the literature. (Author)

  14. BWR type reactor system

    International Nuclear Information System (INIS)

    Morooka, Shin-ichi.

    1980-01-01

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

  15. BWR type reactors

    International Nuclear Information System (INIS)

    Yano, Ryoichi; Sato, Takashi; Osaki, Masahiko; Hirayama, Fumio; Watabe, Atsushi.

    1980-01-01

    Purpose: To effectively eliminate radioactive substances released upon loss of coolant accidents in BWR type reactors. Constitution: A high pressure gas jetting device having a plurality of small aperture nozzles is provided above a spray nozzle, that is, at the top of a dry well. The jetting device is connected to a vacuum breaker provided in a pressure suppression chamber. Upon loss of coolant accident, coolants are sprayed from the spray nozzle and air or nitrogen is jetted from the gas jetting device as well. Then, the gases in the dry well are disturbed, whereby radioactive iodine at high concentration liable to be accumulated in the dry well is forced downwardly, dissolved in the spray water and eliminated. (Ikeda, J.)

  16. Fundamentals of boiling water reactor (BWR)

    International Nuclear Information System (INIS)

    Bozzola, S.

    1982-01-01

    These lectures on fundamentals of BWR reactor physics are a synthesis of known and established concepts. These lectures are intended to be a comprehensive (even though descriptive in nature) presentation, which would give the basis for a fair understanding of power operation, fuel cycle and safety aspects of the boiling water reactor. The fundamentals of BWR reactor physics are oriented to design and operation. In the first lecture general description of BWR is presented, with emphasis on the reactor physics aspects. A survey of methods applied in fuel and core design and operation is presented in the second lecture in order to indicate the main features of the calculational tools. The third and fourth lectures are devoted to review of BWR design bases, reactivity requirements, reactivity and power control, fuel loading patterns. Moreover, operating limits are reviewed, as the actual limits during power operation and constraints for reactor physics analyses (design and operation). The basic elements of core management are also presented. The constraints on control rod movements during the achieving of criticality and low power operation are illustrated in the fifth lecture. Some considerations on plant transient analyses are also presented in the fifth lecture, in order to show the impact between core and fuel performance and plant/system performance. The last (sixth) lecture is devoted to the open vessel testing during the startup of a commercial BWR. A control rod calibration is also illustrated. (author)

  17. BWR type reactors

    International Nuclear Information System (INIS)

    Watanabe, Shoichi

    1986-01-01

    Purpose: To enable to remove water not by way of mechanical operation in a reactor core and improve the fuel economy in BWR type reactors. Constitution: A hollow water removing rod of a cross-like profile made of material having a smaller neutron absorption cross section than the moderator is disposed to the water gap for each of unit structures composed of four fuel assemblies, and water is charged and discharged to and from the water removing rod. Water is removed from the water removing rod to decrease the moderators in the water gap to carry out neutron spectrum shift operation from the initial to the medium stage of reactor core cycles. At the final stage of the cycle, airs in the water removing rod are extracted and the moderator is introduced. The moderator is filled and the criticality is maintained with the accumulated nuclear fission materials. The neutron spectrum shift operation can be attained by eliminating hydrothermodynamic instability and using a water removing rod of a simple structure. (Horiuchi, T.)

  18. BWR type reactor core

    International Nuclear Information System (INIS)

    Tatemichi, Shin-ichiro.

    1981-01-01

    Purpose: To eliminate the variation in the power distribution of a BWR type reactor core in the axial direction even if the flow rate is increased or decreased by providing a difference in the void coefficient between the upper part and the lower parts of the reactor core, and increasing the void coefficient at the lower part of the reactor core. Constitution: The void coefficient of the lower region from the center to the lower part along the axial direction of a nuclear fuel assembly is increased to decrease the dependence on the flow rate of the axial power distribution of the nuclear fuel assembly. That is, a water/fuel ratio is varied, the water in non-boiled region is increased or the neutron spectrum is varied so as to vary the void coefficient. In order to exemplify it, the rate of the internal pellets of the fuel rod of the nuclear fuel assembly or the shape of the channel box is varied. Accordingly, the power does not considerably vary even if the flow rate is altered since the power is varied in the power operation. (Yoshihara, H.)

  19. Recycling systems for BWR type reactors

    International Nuclear Information System (INIS)

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

    1986-01-01

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

  20. BWR type reactors

    International Nuclear Information System (INIS)

    Hayashi, Katsuhisa; Watanabe, Shigeru.

    1983-01-01

    Purpose: To simplify the structure of control rod driving systems, as well as improve the safety and maintainability thereof. Constitution: Control-rod-guide tubes are disposed vertically above the reactor core and control-rod drives are disposed further thereabove, by which the control rods are moved upwardly and downwardly from above the reactor core through the guide tubes. Further, a partitioning cylinder is provided between the inner cirumferential wall at the upper portion of a pressure vessel and the control-rod-guide tubes and a gas-liquid separator is disposed to the space between the partitioning cylinder and the pressure vessel wall, to which steams generated in the reactor core are introduced. In such a structure of the reactor, since all of the control rods are inserted or extracted by the control rod drive system from above the reactor core, if the control rod drives or the likes should fail and accidentally drop the control rods, they exert in the direction of suppressing the nuclear reaction, whereby the safety can be improved. (Sekiya, K.)

  1. BWR type reactors

    International Nuclear Information System (INIS)

    Tsunoyama, Shigeaki; Tanabe, Akira.

    1979-01-01

    Purpose: To provide a main steam pressure shock absorber for reflecting the effect of the pressure propagation to coolants surface in the reactor core. Constitution: An annular shock absorber having near the water level through holes for water level measurement is provided to the gap between the skirt of a steam separator and a pressure vessel. Pressure waves are made the rapid closure of a main steam check valve. If arrived from the dome to the shock absorber, are mostly reflected to the side of the dome and give no substantial effects on the water surface. If the through holes are made small enough, the effects of pressure waves passing through the holes are negligible if they reach the water surface. (Kawakami, Y.)

  2. Analysis of the microstructural evolution of the damage by neutron irradiation in the pressure vessel of a nuclear power reactor BWR

    International Nuclear Information System (INIS)

    Moranchel y R, M.

    2012-01-01

    Nuclear reactor pressure vessel type BWR, installed in Mexico and in many other countries, are made of an alloy of low carbon steel. The American Society for Testing and Materials (Astm) classifies this alloy as A533-B, class 1. Both the vessel and other internal structures are continuously exposed to the neutron flux from the reactions of fission in nuclear fuel. A large number of neutrons reach the vessel and penetrate certain depth depending on their energy. Its penetration in the neutron collides with the nuclei of the atoms out of their positions in the crystal lattice of steel, producing vacancies, interstitial, segregations, among other defects, capable of affecting its mechanical properties. Analyze the micro-structural damage to the vessel due to neutron irradiation, is essential for reasons of integrity of this enclosure and safety of any nuclear power plant. The objective of this thesis work is theoretical and experimentally determine the microstructural damage of a type nuclear reactor vessel steel BWR, due to neutron radiation from the reactor core, using microscopic and spectroscopic techniques as well as Monte Carlo simulation. Microscopy Optical, Scanning Electron Microscopy, Transmission Electron Microscopy, Energy Dispersion of X-rays Spectrometry and X-rays Diffractometry were the techniques used in this research. These techniques helped in the characterization of both the basis of design of pressure vessel steel and steel irradiated, after eight years of neutron irradiation on the vessel, allowing know the surface morphology and crystal structures of the previous steel and post-irradiation, analyze the change in the microstructure of the steel vessel, morphological damage to surface level in an irradiated sample, among which are cavities in the order of microns produced by Atomic displacements due to the impact of neutronic, above all in the first layers of thickness of the vessel, the effect of swelling, regions of greater damage and Atomic

  3. Neutron activation analysis and activity in the vessel steel of a BWR reactor for their study without radiological risks in microscopy and spectrometry

    International Nuclear Information System (INIS)

    Moranchel, M.; Garcia B, A.; Longoria G, L. C.

    2012-01-01

    The vessel material of nuclear reactors is subject to irradiation damage induced by the bombardment of neutrons coming from the reactor core. Neutrons are classified as fast and thermal, which produce different effects. Fast neutrons cause damage to the material by dislocation or displacement of atoms in the crystal structure, while the effect of thermal neutrons is a nuclear transmutation that can significantly change the properties of the material. The type and intensity of damage is based on the characteristics of the material, the flow of neutrons and the modes of neutrons interaction with the atomic structures of the material, among others. This work, alluding to nuclear transmutation, makes an analysis of neutron activation of all isotopes in a steel boiling water nuclear reactor (BWR) vessel. An analytical expression is obtained in order to model activity of steel, on the basis of the weight percentage of its atomic components. Its activity is theoretically estimated in a witness sample of the same material as that of the vessel, placed within the nuclear reactor since the beginning of its commercial operation in April 1995, up to August 2010. It was theoretically determined that the witness sample, with a 0.56 g mass (1 x 1 x 0.07 cm 3 dimensions or equivalent) does not present a radiological risks during the stage of preparation, observation and analysis of it in electron microscopy and X-ray diffraction equipment s. The theoretical results were checked experimentally by measuring the activity of the sample by means of gamma spectrometry, measurement of the exposure levels around the sample, as well as the induced level to whole body and limbs, using thermo-luminescent dosimetry (TLD). As a result of the theoretical analysis, new chemical elements are predicted, as a result of the activation phenomena and radioactive decay, whose presence can be a fundamental factor of change in the properties of the vessel. This work is a preamble to the investigation of

  4. The BWR vessel and internals project - 2001 and beyond

    International Nuclear Information System (INIS)

    Wagoner, V.; Mulford, T.

    2001-01-01

    The BWR Vessel and Internals Project (BWRVIP) is an international association of utilities owning boiling water reactors (BWRs). Figure 1 contains a list of current BWRVIP member utilities. The association was formed in 1994 with the following objectives: to lead the BWR industry toward generic resolution of reactor pressure vessel and internals material condition issues; to identify or develop generic, cost-effective strategies from which each operating plant will select the most appropriate alternative; to serve as the focal point for the regulatory interface with the industry on BWR vessel and internals issues; and to share information and promote communication and cooperation among participating utilities. The initial issue faced by the BWRVIP was core shroud cracking that had been observed in a number of BWRs. The BWRVIP reacted by quickly developing a set of industry guidelines to assist utilities in inspecting, evaluating, and, if necessary, repairing cracked shrouds. Subsequently, the BWRVIP pro-actively developed a comprehensive set of guidelines for managing degradation in other reactor internal components, including the reactor pressure vessel itself. The major components addressed by the BWRVIP are included. (author)

  5. The BWR vessel and internals project - 2001 and beyond

    Energy Technology Data Exchange (ETDEWEB)

    Wagoner, V. [Carolina Power and Light, Progress Energy Building, NC (United States); Mulford, T. [Electric Power Research Institute, Palo Alto, CA (United States)

    2001-07-01

    The BWR Vessel and Internals Project (BWRVIP) is an international association of utilities owning boiling water reactors (BWRs). Figure 1 contains a list of current BWRVIP member utilities. The association was formed in 1994 with the following objectives: to lead the BWR industry toward generic resolution of reactor pressure vessel and internals material condition issues; to identify or develop generic, cost-effective strategies from which each operating plant will select the most appropriate alternative; to serve as the focal point for the regulatory interface with the industry on BWR vessel and internals issues; and to share information and promote communication and cooperation among participating utilities. The initial issue faced by the BWRVIP was core shroud cracking that had been observed in a number of BWRs. The BWRVIP reacted by quickly developing a set of industry guidelines to assist utilities in inspecting, evaluating, and, if necessary, repairing cracked shrouds. Subsequently, the BWRVIP pro-actively developed a comprehensive set of guidelines for managing degradation in other reactor internal components, including the reactor pressure vessel itself. The major components addressed by the BWRVIP are included. (author)

  6. Hydrogen injection device in BWR type reactor

    International Nuclear Information System (INIS)

    Takagi, Jun-ichi; Kubo, Koji.

    1988-01-01

    Purpose: To reduce the increasing ratio of main steam system dose rate due to N-16 activity due to excess hydrogen injection in the hydrogen injection operation of BWR type reactors. Constitution: There are provided a hydrogen injection mechanism for injecting hydrogen into primary coolants of a BWR type reactor, and a chemical injection device for injecting chemicals such as methanol, which makes nitrogen radioisotopes resulted in the reactor water upon hydrogen injection non-volatile, into the pressure vessel separately from hydrogen. Injected hydrogen and the chemicals are not reacted in the feedwater system, but the reaction proceeds due to the presence of radioactive rays after the injection into the pressure vessel. Then, hydrogen causes re-combination in the downcomer portion to reduce the dissolved oxygen concentration. Meanwhile, about 70 % of the chemicals is supplied by means of a jet pump directly to the reactor core, thereby converting the chemical form of N-16 in the reactor core more oxidative (non-volatile). (Kawakami, Y.)

  7. Power oscillations in BWR reactors

    International Nuclear Information System (INIS)

    Espinosa P, G.

    2002-01-01

    One of the main problems in the operation of BWR type reactors is the instability in power that these could present. One type of oscillations and that is the objective of this work is the named density wave, which is attributed to the thermohydraulic processes that take place in the reactor core. From the beginnings of the development of BWR reactors, the stability of these has been an important aspect in their design, due to its possible consequences on the fuel integrity. The reactor core operates in two phase flow conditions and it is observed that under certain power and flow conditions, power instabilities appear. Studying this type of phenomena is complex, due to that a reactor core is constituted approximately by 27,000 fuel bars with different distributions of power and flow. The phenomena that cause the instability in BWR reactors continue being matter of scientific study. In the literature mainly in nuclear subject, it can be observed that exist different methods and approximations for studying this type of phenomena, nevertheless, their results are focused to establish safety limits in the reactor operation, instead of studying in depth of the knowledge about. Also in this line sense of the reactor data analysis, the oscillations characteristic frequencies are obtained for trying to establish if the power is growing or decreasing. In addition to that before mentioned in this paper it is presented a rigorous study applying the volumetric average method, for obtaining the vacuum waves propagation velocities and its possible connection with the power oscillations. (Author)

  8. Water level measurement system in reactor pressure vessel of BWR and hydrogen concentration monitoring system for severe accident

    International Nuclear Information System (INIS)

    Kuroda, Hidehiko; Okazaki, Koki; Shiraishi, Fujio; Kenjyo, Hiroaki; Isoda, Koichiro

    2013-01-01

    TEPCO's Fukushima Daiichi Nuclear Power Station Accident caused severe accident to lose functions of many instrumentation systems. As a result, many important plant parameters couldn't be monitored. In order to monitor plant parameters in the case of severe accident, new instrumentation systems available in the severe conditions are being developed. Water level in reactor pressure vessel and hydrogen concentration in primary containment vessel are one of the most important parameters. Performance test results about water level measurement sensor and hydrogen sensor in severe environmental conditions are described. (author)

  9. BWR type reactors

    International Nuclear Information System (INIS)

    Kikuchi, Osamu.

    1985-01-01

    Purpose: To prevent cavitations in a recycling pump, as well as improve the safety and a reliability of a pressure vessel. Constitution: A feedwater pipeway is connected to the route between the pressure vessel and the recycling pipe and feedwater from the feedwater pipeway is directly introduced to the recycling pump. The temperature of water flowing into the recycling pump is lowered by the feedwater from the feedwater pipeway to prevent the cavitations. (Yoshino, Y.)

  10. BWR type reactor

    International Nuclear Information System (INIS)

    Watanabe, Shoichi

    1983-01-01

    Purpose : To flatten the radial power distribution in the reactor core thereby improve the thermal performance of the reactor core by making the moderator-fuel ratio of fuel assemblies different depending on their position in the reactor core. Constitution : The volume of fuels disposed in the peripheral area of the reactor core is decreased by the increase of the volume of moderators in fuel assemblies disposed in the peripheral area of the reactor core to thereby make the moderator-fuel volume greater in the peripheral area than that in the central area. The moderator-fuel ratio adjustment is attained by making the number of water rods greater, decreasing the diameter of fuel pellets or decreasing the number of fuel pins in fuel assemblies disposed at the peripheral area of the reactor core as compared with fuel assemblies disposed at the central area of the reactor core. In this way, the infinite multiplication factors of fuels can be increased to thereby improve the reactor core performance. (Aizawa, K.)

  11. BWR type reactors

    International Nuclear Information System (INIS)

    Nakajima, Yoshitaka

    1983-01-01

    Purpose: To decrease the control rod exchanging frequency by increasing the working life of control rods for ordinary operation with large neutron irradiation dose, to thereby decrease the exposure dose for operators performing exchanging work, as well as decrease the amount of radioactive wastes resulted upon exchange of the control rods. Constitution: Hafnium solid metal is employed as the neutron absorber of control rods for usual operation inserted into and withdrawn from fuel assemblies for the reactor power control over the entire cycle of the ordinary reactor operation and boron carbide powder is employed as the neutron absorber for emergency control rods to be inserted between the fuel assemblies only upon reactor scram or shutdown, whereby the working life of the control rods for ordinary reactor operation with greater neutron irradiation dose can be improved. Accordingly, the control rod exchanging frequency can be reduced to decrease the exposure dose to the operator for conducting the exchanging work. (Yoshihara, H.)

  12. BWR type reactor

    International Nuclear Information System (INIS)

    Watanabe, Shoichi

    1988-01-01

    Purpose: To inhibit the lowering of the neutron moderation effect due to voids in the upper portion of the reactor core, thereby flatten the axial power distribution. Constitution: Although it has been proposed to enlarge the diameter at the upper portion of a water rod thereby increasing the moderator in the upper portion, since the water rod situates within the channel box, the increase in the capacity thereof is has certain limit. In the present invention, it is designed such that the volume of the region at the outside of the channel box for the fuel assembly to which non-boiling water in the non-boiling water region can enter is made greater in the upper portion than in the lower portion of the reactor core. Thus, if the moderator density in the upper portion of the reactor core should be decreased due to the generation of the voids, the neutron moderating effect in the upper portion of the reactor core is not lowered as compared with the lower portion of the reactor core and, accordingly, the axial power distribution can be flattening more as compared with that in the conventional nuclear reactors. (Takahashi, M.)

  13. Calculation of the neutron flux and fluence in the covering of the nucleus and the vessel of a BWR; Calculo del flujo neutronico y fluencia en la envolvente del nucleo y la vasija de un reactor nuclear BWR

    Energy Technology Data Exchange (ETDEWEB)

    Martinez C, E.; Del Valle G, E. [IPN, Escuela Superior de Fisica y Matematicas, U. P. Adolfo Lopez Mateos, Col. Lindavista, 07738 Mexico D. F. (Mexico); Longoria G, L. C., E-mail: evalle@esfm.ipn.mx [ININ, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico)

    2011-11-15

    One of the main objectives related with the safety in any nuclear power plant, including the nuclear power plant of Laguna Verde, is to guarantee the structural integrity of the pressure vessel of the reactor. To identify and quantifying the damage caused be neutron irradiation in the vessel of any nuclear reactor, is necessary to know as much the neutron flux as the fluence that it has been receiving during their time of operation life, since the observables damages by means of tests mechanics are products of micro-structural effects, induced by neutron irradiation, therefore, is important the study and prediction of the neutron flux to have a better knowledge of the damage that are receiving these materials. In our calculation the code DORT was used, which solves the transport equation in discreet coordinates and in two dimensions (x-y, r-{theta} and r-z), in accord to the regulator guide, it requires to make and approach of the neutron flux in three dimensions by means of the Synthesis Method. Whit this method is possible to achieve a representation of the flux in 3D combining or synthesizing the calculated fluxes by DORT code in r-{theta}, r-z and r. In this work the application of the Synthesis Method is presented, according to the Regulator Guide 1.190, to determine the fluxes 3D in the interns of a BWR using three different space meshes. (Author)

  14. BWR reactor management system

    International Nuclear Information System (INIS)

    Makino, Kakuji; Kawamura, Atsuo; Yoshioka, Ritsuo; Neda, Toshikatsu.

    1979-01-01

    It is necessary to grasp the delicate state of operation in reactor cores in view of the control of burn-up and power output at the time of the operation management of BWRs. Enormous labor has been required for the collection, processing and evaluation of the data. It is desirable to obtain the safer, more efficient and faster method of operation control by predicting the states in cores including the change of xenon and reflecting them to operation plans as well as by tracing with high accuracy the past burn-up history for a long period. At present, the on-line evaluation of the states in cores is carried out with the process computers attached to respective units, but the amount of data required for core operation management of high degree far exceeds their capacity. From such viewpoints, the research and development on the reactor management system were carried out. The data processing concerning core operation management is performed with newly installed computers utilizing the data from existing process computers, and the operation of reactor cores, the qualitative improvement of management works, labor saving, and fast, efficient operation control are feasible with it. This system was installed in an actual plant in October, 1977. The composition of the system, the prediction of the change in local output distribution accompanying control rod operation, the prediction of the change in the states in cores due to the flow rate of coolant, and the function of collecting plant data are explained. (Kako, I.)

  15. NDE and fracture mechanics evaluation of bottom-head weld indications in a BWR reactor pressure vessel

    Energy Technology Data Exchange (ETDEWEB)

    Brickstad, B [Swedish Plant Inspectorate, Stockholm (Sweden)

    1988-12-31

    This document deals with the Non Destructive Examination (NDE) and the fracture mechanics evaluation of bottom head welds in a BWR. The NDE equipment is presented, together with the geometry of evaluated flaw regions. After the fracture mechanics evaluation, it appeared that the plant results fulfilled the usual conditions, and the plant was allowed to operate one more year. (TEC).

  16. Pressurized Water Reactors (PWR) and Boiling Water Reactors (BWR) are compared

    International Nuclear Information System (INIS)

    Greneche, D.

    2014-01-01

    This article compares the 2 types of light water reactors that are used to produce electricity: the Pressurized Water Reactor (PWR) and the Boiling Water Reactor (BWR). Historically the BWR concept was developed after the PWR concept. Today 80% of light water reactors operating in the world are of PWR-type. This comparison is comprehensive and detailed. First the main technical features are reviewed and compared: reactor architecture, core and fuel design, reactivity control, reactor vessel, cooling systems and reactor containment. Secondly, various aspects concerning reactor operations like reactor control, fuel management, maintenance, inspections, radiation protection, waste generation and reactor reliability are presented and compared for both reactors. As for the issue of safety, it is highlighted that the accidental situations are too different for the 2 reactors to be compared. The main features of reactor safety are explained for both reactors

  17. Reliable estimation of neutron flux in BWR reactor vessel using the tort code (2) application to neutron and gamma flux estimation

    Energy Technology Data Exchange (ETDEWEB)

    Kurosawa, M. [Toshiba Corp., Yokohama (Japan); Tsukiyama, T.; Hayashi, K. [Hitachi Engineering Co. Ltd., Hitachi-shi (Japan)

    2001-07-01

    A neutron and gamma flux distribution around the core of BWR commercial plant in Japan was calculated, using a three-dimensional transport code, TORT in DOORS32 code system. In the external of the core, the bottom of the model was at an elevation of 150 cm below the bottom of active fuel, the top of the model was at an elevation of the top of the shroud head dome and the radial part of the model was to the outside of the reactor pressure vessel. The top guide beams were modeled explicitly to obtain the neutron and gamma flux distribution both in the beams and outside beams. The each control rod guide tube was also modeled with homogeneous region which included the blade wing and poison tubes so that we could obtain the neutron and gamma flux distribution around the each control rod guide tube. The calculation model mentioned above needed very large memory size which exceeded a few decade giga-bytes. As the using the splicing/coupling method had uncertainly at the splicing/coupling boundary, in this work the calculation was performed without this splicing/coupling method. On the other hand, radioactivity data were measured for a few pieces of the top guide beam, shroud and in-core monitor guide tube in the same plant which was analyzed in the above calculation. So the calculation results were able to be compared with those measured data as benchmarking and at the end of this task, the C/M values at these measured points were obtained and calculation model using TORT was evaluated. (authors)

  18. Analysis of the microstructural evolution of the damage by neutron irradiation in the pressure vessel of a nuclear power reactor BWR; Analisis de la evolucion microestructural del dano por irradiacion neutronica en la vasija de presion de un reactor nuclear de potencia BWR

    Energy Technology Data Exchange (ETDEWEB)

    Moranchel y R, M.

    2012-07-01

    Nuclear reactor pressure vessel type BWR, installed in Mexico and in many other countries, are made of an alloy of low carbon steel. The American Society for Testing and Materials (Astm) classifies this alloy as A533-B, class 1. Both the vessel and other internal structures are continuously exposed to the neutron flux from the reactions of fission in nuclear fuel. A large number of neutrons reach the vessel and penetrate certain depth depending on their energy. Its penetration in the neutron collides with the nuclei of the atoms out of their positions in the crystal lattice of steel, producing vacancies, interstitial, segregations, among other defects, capable of affecting its mechanical properties. Analyze the micro-structural damage to the vessel due to neutron irradiation, is essential for reasons of integrity of this enclosure and safety of any nuclear power plant. The objective of this thesis work is theoretical and experimentally determine the microstructural damage of a type nuclear reactor vessel steel BWR, due to neutron radiation from the reactor core, using microscopic and spectroscopic techniques as well as Monte Carlo simulation. Microscopy Optical, Scanning Electron Microscopy, Transmission Electron Microscopy, Energy Dispersion of X-rays Spectrometry and X-rays Diffractometry were the techniques used in this research. These techniques helped in the characterization of both the basis of design of pressure vessel steel and steel irradiated, after eight years of neutron irradiation on the vessel, allowing know the surface morphology and crystal structures of the previous steel and post-irradiation, analyze the change in the microstructure of the steel vessel, morphological damage to surface level in an irradiated sample, among which are cavities in the order of microns produced by Atomic displacements due to the impact of neutronic, above all in the first layers of thickness of the vessel, the effect of swelling, regions of greater damage and Atomic

  19. Neutron activation analysis and activity in the vessel steel of a BWR reactor for their study without radiological risks in microscopy and spectrometry; Analisis de activacion neutronica y actividad en el acero de la vasija de un reactor nuclear tipo BWR para su estudio sin riesgos radiologicos en microscopia y espectrometria

    Energy Technology Data Exchange (ETDEWEB)

    Moranchel, M.; Garcia B, A. [IPN, Escuela Superior de Fisica y Matematicas, Departamento de Fisica, Unidad Profesional Adolfo Lopez Mateos, Zacatenco, 07738 Mexico D. F. (Mexico); Longoria G, L. C., E-mail: mmoranchel@ipn.mx [IAEA, Department of Technical Cooperation, Division for Latin America, Room B1109 Wagramerstrasse 5, PO Box 100, A-1400, Vienna (Austria)

    2012-07-01

    The vessel material of nuclear reactors is subject to irradiation damage induced by the bombardment of neutrons coming from the reactor core. Neutrons are classified as fast and thermal, which produce different effects. Fast neutrons cause damage to the material by dislocation or displacement of atoms in the crystal structure, while the effect of thermal neutrons is a nuclear transmutation that can significantly change the properties of the material. The type and intensity of damage is based on the characteristics of the material, the flow of neutrons and the modes of neutrons interaction with the atomic structures of the material, among others. This work, alluding to nuclear transmutation, makes an analysis of neutron activation of all isotopes in a steel boiling water nuclear reactor (BWR) vessel. An analytical expression is obtained in order to model activity of steel, on the basis of the weight percentage of its atomic components. Its activity is theoretically estimated in a witness sample of the same material as that of the vessel, placed within the nuclear reactor since the beginning of its commercial operation in April 1995, up to August 2010. It was theoretically determined that the witness sample, with a 0.56 g mass (1 x 1 x 0.07 cm{sup 3} dimensions or equivalent) does not present a radiological risks during the stage of preparation, observation and analysis of it in electron microscopy and X-ray diffraction equipment s. The theoretical results were checked experimentally by measuring the activity of the sample by means of gamma spectrometry, measurement of the exposure levels around the sample, as well as the induced level to whole body and limbs, using thermo-luminescent dosimetry (TLD). As a result of the theoretical analysis, new chemical elements are predicted, as a result of the activation phenomena and radioactive decay, whose presence can be a fundamental factor of change in the properties of the vessel. This work is a preamble to the

  20. Primary coolant system of BWR type reactor

    International Nuclear Information System (INIS)

    Ibe, Hidefumi; Takahashi, Masanori; Aoki, Yasuko

    1997-01-01

    The present invention provides a water quality control system for preventing corrosion and for extending working life of structural materials of a BWR-type reactor. Namely, a sensor group 1 and a sensor group 2 are disposed at different positions such as in a feedwater system, a recycling system, main steam pipes, and a pressure vessel, respectively. Each sensor group can record and generate alarms independently. The sensor group 1 for usual monitoring is connected to a calculation device by way of a switch to confirm that the monitored values are within a proper range by the injection of a water quality moderating agent. The sensor group 2 is caused to stand alone or connected with the calculation device by way of a switch optionally. When abnormality should occur in the sensor group 1, the sensor group 2 determines the limit for the increase/decrease of controlling amount of the moderating agent at a portion where the conditions are changed to the most severe direction by using data base. The moderating agent is injected and controlled based on the controlling amount. The system of the present invention can optionally cope with a new sensor and determination for new water quality standards. Then the evaluation/control accuracy of the entire system can be improved while covering up the errors of each sensor. (I.S.)

  1. Inservice inspection of Halden BWR pressure vessel

    International Nuclear Information System (INIS)

    Foerli, O.; Hernes, T.

    1978-01-01

    A description is given of how the recertification inspection of the 20 years old Halden Reactor pressure vessel was carried out in accordance with the latest ASME-CODES, despite the fact that inspection accessibility was poor. As no volumetric inspection had been carried out since the preservice radiography in 1957, the ultrasonic inspection included the high flux region of all welds. In total 70% of longitudinal welds and 20% of bottom circumferential welds were inspected as well as the bottom nozzle connection. The vessel was not designed with provisions for inservice inspection, the welds are unaccessible from the outside and removal of the lid is virtually impossible. The ultrasonic probes could only be loaded through 77 mm diameter holes in the top lid and remotely positioned inside the vessel. The inspection was performed using 450C and 60OC 1 MHz angle probes and 2.25 MHz normal probes in immersion technique. In a zone around the welds, small regions with lack of bonding between the stainless steel cladding and the boiler steel were revealed. One root defect known and accepted from the preservice radiographs was examined. The defect was found to be 6x30mm as a maximum and well within acceptable limits according to the fracture mechanics analysis method recommended in ASME X1. The inspection required a period of three weeks' work in the reactor hall. (UK)

  2. Identification and assessment of BWR in-vessel severe accident mitigation strategies

    International Nuclear Information System (INIS)

    Hodge, S.A.; Kress, T.S.; Cleveland, J.C.; Petek, M.

    1992-01-01

    This paper briefly describes the results of work carried out in support of the US Nuclear Regulatory Commission Accident Management Research Program to evaluate the effectiveness and feasibility of current and proposed strategies for BWR severe accident management. These results are described in detail in the just-released report Identification and Assessment of BWR In-Vessel Severe Accident Mitigation Strategies, NUREG/CR-5869, which comprises three categories of findings. First, an assessment of the current status of accident management strategies for the mitigation of in-vessel events for BWR severe accident sequences is combined with a review of the BWR Owners' Group Emergency Procedure Guidelines (EPGs) to determine the extent to which they currently address the characteristic events of an unmitigated severe accident. Second, where considered necessary, new candidate accident management strategies are proposed for mitigation of the late-phase (after core damage has occurred) events. Finally, two of the four candidate strategies identified by this effort are assessed in detail. These are (1) preparation of a boron solution for reactor vessel refill should control blade damage occur during a period of temporary core dryout and (2) containment flooding to maintain the core debris within the reactor vessel if the injection systems cannot be restored

  3. Power control system in BWR type reactors

    International Nuclear Information System (INIS)

    Nishizawa, Yasuo.

    1980-01-01

    Purpose: To control the reactor power so that the power distribution can satisfy the limiting conditions, by regulating the reactor core flow rate while monitoring the power distribution in the reactor core of a BWR type reactor. Constitution: A power distribution monitor determines the power distribution for the entire reactor core based on the data for neutron flux, reactor core thermal power, reactor core flow rate and control rod pattern from the reactor and calculates the linear power density distribution. A power up ratio computing device computes the current linear power density increase ratio. An aimed power up ratio is determined by converting the electrical power up ratio transferred from a load demand input device into the reactor core thermal power up ratio. The present reactor core thermal power up ratio is subtracted from the limiting power up ratio and the difference is sent to an operation amount indicator and the reactor core flow rate is changed in a reactor core flow rate regulator, by which the reactor power is controlled. (Moriyama, K.)

  4. Feedwater control system in BWR type reactor

    International Nuclear Information System (INIS)

    Tanji, Jun-ichi; Oomori, Takashi.

    1980-01-01

    Purpose: To improve the water level control performance in BWR type reactor by regulating the water level set to the reactor depending on the rate of change in the recycling amount of coolant to thereby control the fluctuations in the water level resulted in the reactor within an aimed range even upon significant fluctuations in the recycling flow rate. Constitution: The recycling flow rate of coolant in the reactor is detected and the rate of its change with time is computed to form a rate of change signal. The rate of change signal is inputted to a reactor level setter to amend the actual reactor water level demand signal and regulate the water level set to the reactor water depending on the rate of change in the recycling flow rate. Such a regulation method for the set water level enables to control the water level fluctuation resulted in the reactor within the aimed range even upon the significant fluctuation in the recycling flow rate and improve the water level control performance of the reactor, whereby the operationability for the reactor is improved to enhance the operation rate. (Moriyama, K.)

  5. Recent BWR fuel management reactor physics advances

    International Nuclear Information System (INIS)

    Crowther, R.L.; Congdon, S.P.; Crawford, B.W.; Kang, C.M.; Martin, C.L.; Reese, A.P.; Savoia, P.J.; Specker, S.R.; Welchly, R.

    1982-01-01

    Improvements in BWR fuel management have been under development to reduce uranium and separative work (SWU) requirements and reduce fuel cycle costs, while also maintaining maximal capacity factors and high fuel reliability. Improved reactor physics methods are playing an increasingly important role in making such advances feasible. The improved design, process computer and analysis methods both increase knowledge of the thermal margins which are available to implement fuel management advance, and improve the capability to reliably and efficiently analyze and design for fuel management advances. Gamma scan measurements of the power distributions of advanced fuel assembly and advanced reactor core designs, and improved in-core instruments also are important contributors to improving 3-d predictive methods and to increasing thermal margins. This paper is an overview of the recent advances in BWR reactor physics fuel management methods, coupled with fuel management and core design advances. The reactor physics measurements which are required to confirm the predictions of performance fo fuel management advances also are summarized

  6. Panorama of the BWR reactors - Evolution of the concept

    Energy Technology Data Exchange (ETDEWEB)

    Novotny, C.; Uhrig, E. [AREVA NP GmbH, Safety Engineering Department - PEPS-G (Germany)

    2012-01-15

    Nowadays, a fleet of more than 50 boiling water reactors (BWR) are in operation in the world. This article gives a short overview on the developments of nuclear power plants of the BWR type, with a focus on the European builds. It describes the technical bases from the early designs in the fifties, sketches the innovations of the sixties and seventies in the types BWR 69 and 72 (Baulinie 69 and 72) and gives an outlook of a possible next generation BWR. A promising approach in recent BWR developments is the the combination of passive safety systems with established design basis

  7. Strategies of operation cycles in BWR type reactors

    International Nuclear Information System (INIS)

    Molina, D.; Sendino, F.

    1996-01-01

    The article analyzes the operation cycles in BWR type reactors. The cycle size of operation is the consequence on the optimization process of the costs with the technical characteristics of nuclear fuel and the characteristics of demand and production. The authors analyze the cases of Garona NP and Cofrentes NP, both with BWR reactors. (Author)

  8. Level controlling system in BWR type reactors

    International Nuclear Information System (INIS)

    Joge, Toshio; Higashigawa, Yuichi; Oomori, Takashi.

    1981-01-01

    Purpose: To reasonably attain fully automatic water level control in the core of BWR type nuclear power plants. Constitution: A feedwater flow regulation valve for reactor operation and a feedwater flow regulation valve for starting are provided at the outlet of a motor-driven feedwater pump in a feedwater system, and these valves are controlled by a feedwater flow rate controller. While on the other hand, a damp valve for reactor clean up system is controlled either in ''computer'' mode or in ''manual'' mode selected by a master switch, that is, controlled from a computer or the ON-OFF switch of the master switch by way of a valve control analog memory and a turn-over switch. In this way, the water level in the nuclear reactor can be controlled in a fully automatic manner reasonably at the starting up and shutdown of the plant to thereby provide man power saving. (Seki, T.)

  9. Identification and assessment of BWR in-vessel severe accident mitigation strategies

    International Nuclear Information System (INIS)

    Hodge, S.A.; Cleveland, J.C.; Kress, T.S.; Petek, M.

    1992-10-01

    This report provides the results of work carried out in support of the US Nuclear Regulatory Commission Accident Management Research Program to develop a technical basis for evaluating the effectiveness and feasibility of current and proposed strategies for boiling water reactor (BWR) severe accident management. First, the findings of an assessment of the current status of accident management strategies for the mitigation of in-vessel events for BWR severe accident sequences are described. This includes a review of the BWR Owners' Group Emergency Procedure Guidelines (EPGSs) to determine the extent to which they currently address the characteristic events of an unmitigated severe accident and to provide the basis for recommendations for enhancement of accident management procedures. Second, where considered necessary, new candidate accident management strategies are proposed for mitigation of the late-phase (after core damage has occurred) events. Finally, recommendations are made for consideration of additional strategies where warranted, and two of the four candidate strategies identified by this effort are assessed in detail: (1) preparation of a boron solution for reactor vessel refill should control blade damage occur during a period of temporary core dryout and (2) containment flooding to maintain the core debris within the reactor vessel if the injection systems cannot be restored

  10. Identification and assessment of BWR in-vessel severe accident mitigation strategies

    Energy Technology Data Exchange (ETDEWEB)

    Hodge, S.A.; Cleveland, J.C.; Kress, T.S.; Petek, M. [Oak Ridge National Lab., TN (United States)

    1992-10-01

    This report provides the results of work carried out in support of the US Nuclear Regulatory Commission Accident Management Research Program to develop a technical basis for evaluating the effectiveness and feasibility of current and proposed strategies for boiling water reactor (BWR) severe accident management. First, the findings of an assessment of the current status of accident management strategies for the mitigation of in-vessel events for BWR severe accident sequences are described. This includes a review of the BWR Owners` Group Emergency Procedure Guidelines (EPGSs) to determine the extent to which they currently address the characteristic events of an unmitigated severe accident and to provide the basis for recommendations for enhancement of accident management procedures. Second, where considered necessary, new candidate accident management strategies are proposed for mitigation of the late-phase (after core damage has occurred) events. Finally, recommendations are made for consideration of additional strategies where warranted, and two of the four candidate strategies identified by this effort are assessed in detail: (1) preparation of a boron solution for reactor vessel refill should control blade damage occur during a period of temporary core dryout and (2) containment flooding to maintain the core debris within the reactor vessel if the injection systems cannot be restored.

  11. Reactor vessel sealing plug

    International Nuclear Information System (INIS)

    Dooley, R.A.

    1986-01-01

    This invention relates to an apparatus and method for sealing the cold leg nozzles of a nuclear reactor pressure vessel from a remote location during maintenance and inspection of associated steam generators and pumps while the pressure vessel and refueling canal are filled with water. The apparatus includes a sealing plug for mechanically sealing the cold leg nozzle from the inside of a reactor pressure vessel. The sealing plugs include a primary and a secondary O-ring. An installation tool is suspended within the reactor vessel and carries the sealing plug. The tool telescopes to insert the sealing plug within the cold leg nozzle, and to subsequently remove the plug. Hydraulic means are used to activate the sealing plug, and support means serve to suspend the installation tool within the reactor vessel during installation and removal of the sealing plug

  12. Method of operating BWR type reactors

    International Nuclear Information System (INIS)

    Sekimizu, Koichi

    1980-01-01

    Purpose: To enable reactor control depending on any demanded loads by performing control by the insertion of control rods in addition to the control by the regulation of the flow rate of the reactor core water at high power operation of a BWR type reactor. Method: The power is reduced at high power operation by decreasing the flow rate of reactor core water from the starting time for the power reduction and the flow rate is maintained after the time at which it reaches the minimum allowable flow rate. Then, the control rod is started to insert from the above time point to reduce the power to an aimed level. Thus, the insufficiency in the reactivity due to the increase in the xenon concentration can be compensated by the withdrawal of the control rods and the excess reactivity due to the decrease in the xenon concentration can be compensated by the insertion of the control rods, whereby the reactor power can be controlled depending on any demanded loads without deviating from the upper or lower limit for the flow rate of the reactor core water. (Moriyama, K.)

  13. Identification and initial assessment of candidate BWR late-phase in-vessel accident management strategies

    International Nuclear Information System (INIS)

    Hodge, S.A.

    1991-01-01

    Work sponsored by the United States Nuclear Regulatory Commission (USNRC) to identify and perform preliminary assessments of candidate BWR [boiling water reactor] in-vessel accident management strategies was completed at Oak Ridge National Laboratory (ORNL) during fiscal year 1990. Mitigative strategies for containment events have been the subject of a companion study at Brookhaven National Laboratory. The focus of this Oak Ridge effort was the development of new strategies for mitigation of the late phase events, that is, the events that would occur in-vessel after the onset of significant core damage. The work began with an investigation of the current status of BWR in-vessel accident management procedures and proceeded through a preliminary evaluation of several candidate new strategies. The steps leading to the identification of the candidate strategies are described. The four new candidate late-phase (in-vessel) accident mitigation strategies identified by this study and discussed in the report are: (1) keep the reactor vessel depressurized; (2) restore injection in a controlled manner; (3) inject boron if control blade damage has occurred; and (4) containment flooding to maintain core and structural debris in-vessel. Additional assessments of these strategies are proposed

  14. 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.)

  15. Power generator in BWR type reactors

    International Nuclear Information System (INIS)

    Yoshida, Kenji.

    1984-01-01

    Purpose: To enable to perform stable and dynamic conditioning operation for nuclear fuels in BWR type reactors. Constitution: The conditioning operation for the nuclear fuels is performed by varying the reactor core thermal power in a predetermined pattern by changing the predetermined power changing pattern of generator power, the rising rate of the reactor core thermal power and the upper limit for the rising power of the reactor core thermal power are calculated and the power pattern for the generator is corrected by a power conditioning device such that the upper limit for the thermal power rising rate and the upper limit for the thermal power rising rate are at the predetermined levels. Thus, when the relation between the reactor core thermal power and the generator electrical power is fluctuated, the fluctuation is detected based on the variation in the thermal power rising rate and the limit value for the thermal power rising rate, and the correction is made to the generator power changing pattern so that these values take the predetermined values to thereby perform the stable conditioning operation for the nuclear fuels. (Moriyama, K.)

  16. 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

  17. Fuel assemblies for BWR type reactors

    International Nuclear Information System (INIS)

    Ishizuka, Takao.

    1981-01-01

    Purpose: To enable effective failed fuel detection by the provision of water rod formed with a connecting section connected to a warmed water feed pipe of a sipping device at the lower portion and with a warmed water jetting port in the lower portion in a fuel assembly of a BWR type reactor to thereby carry out rapid sipping. Constitution: Fuel rods and water rods are contained in the channel box of a fuel assembly, and the water rod is provided at its upper portion with a connecting section connected to the warmed water feed pipe of the sipping device and formed at its lower portion with a warmed water jetting port for jetting warmed water fed from the warmed water feed pipe. Upon detection of failed fuels, the reactor operation is shut down and the reactor core is immersed in water. The cover for the reactor container is removed and the cap of the sipping device is inserted to connect the warmed water feed pipe to the connecting section of the water rod. Then, warmed water is fed to the water rod and jetted out from the warmed water jetting port to cause convection and unify the water of the channel box in a short time. Thereafter, specimen is sampled and analyzed for the detection of failed fuels. (Moriyama, K.)

  18. FFTF and CRBRP reactor vessels

    International Nuclear Information System (INIS)

    Morgan, R.E.

    1977-01-01

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

  19. 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)

  20. 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)

  1. Thermalydraulic processes in the reactor coolant system of a BWR under severe accident conditions

    International Nuclear Information System (INIS)

    Hodge, S.A.

    1990-01-01

    Boiling water reactors (BWRs) incorporate many unique structural features that make their expected response under severe accident conditions very different from that predicted in the case of pressurized water reactor accident sequences. Automatic main steam isolation valve (MIV) closure as the vessel water level approaches the top of the core would cause reactor vessel isolation while automatic recirculation pump trip would limit the in-vessel flows to those characteristic of natural circulation (as disturbed by vessel relief valve actuation). This paper provides a discussion of the BWR control blade, channel box, core plate, control rod guide tube, and reactor vessel safety relief valve (SRV) configuration and the effects of these structural components upon thermal hydraulic processes within the reactor vessel under severe accident conditions. The dominant BWR severe accident sequences as determined by probabilistic risk assessment are described and the expected timing of events for the unmitigated short-term station blackout severe accident sequence at the Peach Bottom atomic power station is presented

  2. 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.)

  3. Lightweight submersed 'Walking' NDE manipulators for PWR and BWR vessel weld inspection

    International Nuclear Information System (INIS)

    Saernmark, Ivan; Lenz, Herbert

    2008-01-01

    Three new manipulators developed by WesDyne TRC in Sweden have under the year 2007 performed three very successful inspections in the PWR reactor Ringhals 3 and the BWR reactors Ringhals 1 and Oskarshamn 1. The manipulator systems can be used to perform inspection of circumferential and vertical welds on the reactor pressure vessel, the core shroud, core shroud support in BWR reactors or vessel and core barrel welds in PWR reactors. Most other flat or curved surfaces can be inspected using the new concept through relatively simple mechanical reconfigurations of system modules. The first inspection was performed on the R3 PWR core barrel in June 2007 with a very good result. This Manipulator is designed for access in very narrow gaps and for the type of core barrels with a shield covering the whole area of the perimeter. The manipulator is attached to the inspection area by means of a new unique suction cup system. The current manipulators consist of a curved horizontal beam, with radius similar to the reactor vessel, and a straight vertical beam, forming a T-shaped structure. By alternating the application of suction cup pairs on the horizontal beam and the vertical beam and by driving the scanning motors, the manipulator performs an incremental translational movement upwards/downwards or from side to side. The principles of this system give a well defined and stable platform for global and local positioning accuracy. A combination of advanced sensor solutions provides accurate position information in the absence of other physical reference objects. The system is controlled by the new WesDyne TRC Motor Control Panel and software, the MCP is specifically designed for remote control of submersed manipulators using techniques for cable reduction

  4. Lightweight submersed 'Walking' NDE manipulators for PWR and BWR vessel weld inspection

    Energy Technology Data Exchange (ETDEWEB)

    Saernmark, Ivan; Lenz, Herbert [WesDyne TRC AB, Stockholm (Sweden)

    2008-04-15

    Three new manipulators developed by WesDyne TRC in Sweden have under the year 2007 performed three very successful inspections in the PWR reactor Ringhals 3 and the BWR reactors Ringhals 1 and Oskarshamn 1. The manipulator systems can be used to perform inspection of circumferential and vertical welds on the reactor pressure vessel, the core shroud, core shroud support in BWR reactors or vessel and core barrel welds in PWR reactors. Most other flat or curved surfaces can be inspected using the new concept through relatively simple mechanical reconfigurations of system modules. The first inspection was performed on the R3 PWR core barrel in June 2007 with a very good result. This Manipulator is designed for access in very narrow gaps and for the type of core barrels with a shield covering the whole area of the perimeter. The manipulator is attached to the inspection area by means of a new unique suction cup system. The current manipulators consist of a curved horizontal beam, with radius similar to the reactor vessel, and a straight vertical beam, forming a T-shaped structure. By alternating the application of suction cup pairs on the horizontal beam and the vertical beam and by driving the scanning motors, the manipulator performs an incremental translational movement upwards/downwards or from side to side. The principles of this system give a well defined and stable platform for global and local positioning accuracy. A combination of advanced sensor solutions provides accurate position information in the absence of other physical reference objects. The system is controlled by the new WesDyne TRC Motor Control Panel and software, the MCP is specifically designed for remote control of submersed manipulators using techniques for cable reduction.

  5. Coolant cleanup system for BWR type reactor

    International Nuclear Information System (INIS)

    Kinoshita, Shoichiro; Araki, Hidefumi.

    1993-01-01

    The cleanup system of the present invention removes impurity ions and floating materials accumulated in a reactor during evaporation of coolants in the nuclear reactor. That is, coolants pass pipelines from a pressure vessel using pressure difference between a high pressure in the pressure vessel and a low pressure at the upstream of a condensate filtration/desalting device of a condensate/feed water system as a driving source, during which cations and floating materials are removed in a high temperature filtration/desalting device and coolants flow into the condensate/feedwater system. Impurities containing anions are removed here by the condensates filtration/desalting device. Then, they return to the pressure vessel while pressurized and heated by a condensate pump, a feed water pump and a feed water heater. At least pumps, a heat exchanger for heating, a filtration/desalting device for removing anions and pipelines connecting them used exclusively for the coolant cleanup system are no more necessary. (I.S.)

  6. Simulation of decreasing reactor power level with BWR simulator

    International Nuclear Information System (INIS)

    Suwoto; Zuhair; Rivai, Abu Khalid

    2002-01-01

    Study on characteristic of BWR using Desktop PC Based Simulator Program was analysed. This simulator is more efficient and cheaper for analyzing of characteristic and dynamic respond than full scope simulator for decreasing power level of BW. Dynamic responses of BWR reactor was investigated during the power level reduction from 100% FP (Full Power) which is 3926 MWth to 0% FP with 25% steps and 1 % FP/sec rate. The overall results for core flow rate, reactor steam flow, feed-water flow and turbine-generator power show tendency proportional to reduction of reactor power. This results show that reactor power control in BWR could be done by control of re-circulation flow that alter the density of water used as coolant and moderator. Decreasing the re-circulation flow rate will decrease void density which has negative reactivity and also affect the position of control rods

  7. Fast neutron fluence calculations as support for a BWR pressure vessel and internals surveillance program

    International Nuclear Information System (INIS)

    Lucatero, Marco A.; Palacios-Hernandez, Javier C.; Ortiz-Villafuerte, Javier; Xolocostli-Munguia, J. Vicente; Gomez-Torres, Armando M.

    2010-01-01

    Materials surveillance programs are required to detect and prevent degradation of safety-related structures and components of a nuclear power reactor. In this work, following the directions in the Regulatory Guide 1.190, a calculational methodology is implemented as additional support for a reactor pressure vessel and internals surveillance program for a BWR. The choice of the neutronic methods employed was based on the premise of being able of performing all the expected future survey calculations in relatively short times, but without compromising accuracy. First, a geometrical model of a typical BWR was developed, from the core to the primary containment, including jet pumps and all other structures. The methodology uses the Synthesis Method to compute the three-dimensional neutron flux distribution. In the methodology, the code CORE-MASTER-PRESTO is used as the three-dimensional core simulator; SCALE is used to generate the fine-group flux spectra of the components of the model and also used to generate a 47 energy-groups job cross section library, collapsed from the 199-fine-group master library VITAMIN-B6; ORIGEN2 was used to compute the isotopic densities of uranium and plutonium; and, finally, DORT was used to calculate the two-dimensional and one-dimensional neutron flux distributions required to compute the synthesized three-dimensional neutron flux. Then, the calculation of fast neutron fluence was performed using the effective full power time periods through six operational fuel cycles of two BWR Units and until the 13th cycle for Unit 1. The results showed a maximum relative difference between the calculated-by-synthesis fast neutron fluxes and fluences and those measured by Fe, Cu and Ni dosimeters less than 7%. The dosimeters were originally located adjacent to the pressure vessel wall, as part of the surveillance program. Results from the computations of peak fast fluence on pressure vessel wall and specific weld locations on the core shroud are

  8. 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

  9. 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

  10. 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.

  11. Feedwater recycling system in BWR type reactor

    International Nuclear Information System (INIS)

    Shimamoto, Yoshiharu.

    1980-01-01

    Purpose: To improve the reactor safety by preventing thermal stresses and cracks generated in structural materials due to the fluctuations in the temperature for high temperature water - low temperature water mixture near the feedwater nozzle. Method: Feedwater pipes are connected to a pressure vessel not directly but by way of a flow control valve. While the recycled water is circulated from an inlet nozzle to an outlet nozzle through a recycle pump, flow control valve and recycling pipeways, feedwater is fed from the feedwater pipes to the recycling pipeways by way of the flow control valve. More specifically, since the high temperature recycle water and the low temperature recycle water are mixed within the pipeways, the temperature fluctuations resulted from the temperature difference between the recycle water and the feedwater is reduced to prevent thermal fatigue and generation of cracks thereby securing the reactor safety. (Furukawa, Y.)

  12. Nuclear reactor vessel inspection apparatus

    International Nuclear Information System (INIS)

    Blackstone, E.G.; Lofy, R.A.; Williams, L.P.

    1979-01-01

    Apparatus for the in situ inspection of a nuclear reactor vessel to detect the location and character of flaws in the walls of the vessel, in the welds joining the various sections of the vessel, in the welds joining attachments such as nozzles, elbows and the like to the reactor vessel and in such attachments wherein an inspection head carrying one or more ultrasonic transducers follows predetermined paths in scanning the various reactor sections, welds and attachments

  13. Mobile nuclear reactor containment vessel

    International Nuclear Information System (INIS)

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

    1978-01-01

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

  14. 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.)

  15. Neutron dosimetry. Environmental monitoring in a BWR type reactor

    International Nuclear Information System (INIS)

    Tavera D, L.; Camacho L, M.E.

    1991-01-01

    The measurements carried out on reactor dosimetry are applied mainly to the study on the effects of the radiation in 108 materials of the reactor; little is on the environmental dosimetry outside of the primary container of BWR reactors. In this work the application of a neutron spectrometer formed by plastic detectors of nuclear traces manufactured in the ININ, for the environmental monitoring in penetrations around the primary container of the unit I of the Laguna Verde central is presented. The neutron monitoring carries out with purposes of radiological protection, during the operational tests of the reactor. (Author)

  16. Reactor pressure vessel embrittlement

    International Nuclear Information System (INIS)

    1992-07-01

    Within the framework of the IAEA extrabudgetary programme on the Safety of WWER-440/230 NPPs, a list of safety issues requiring broad studies of generic interest have been agreed upon by an Advisory Group who met in Vienna in September 1990. The list was later revised in the light of the programme findings. The information on the status of the issues, and on the amount of work already completed and under way in the various countries, needs to be compiled. Moreover, an evaluation of what further work is required to resolve each one of the issues is also necessary. In view of this, the IAEA has started the preparation of a series of status reports on the various issues. This report on the generic safety issue ''Reactor Pressure Vessel Embrittlement'' presents a comprehensive survey of technical information available in the field and identifies those aspects which require further investigation. 39 refs, 21 figs, 4 tabs

  17. 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.)

  18. 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

  19. Paired replacement fuel assemblies for BWR-type reactor

    International Nuclear Information System (INIS)

    Oguchi, Kazushige.

    1997-01-01

    There are disposed a large-diameter water rod constituting a non-boiling region at a central portion and paired replacement fuel assemblies for two streams having the same average enrichment degree and different amount of burnable poisons. The paired replacement fuel assemblies comprise a first fuel assembly having a less amount of burnable poisons and a second fuel assembly having a larger amount of burnable poisons. A number of burnable poison-containing fuel rods in adjacent with the large diameter water rod is increased in the second fuel assembly than the first fuel assembly. Then, the poison of the paired replacement fuel assemblies for the BWR type reactor can be annihilated simultaneously at the final stage of the cycle. Accordingly, fuels for a BWR type reactor excellent in economical property and safety and facilitating the design of the replacement reactor core can be obtained. (N.H.)

  20. Neutron dosimetry. Environmental monitoring in a BWR type reactor; Dosimetria de neutrones. Monitoreo ambiental en un reactor del tipo BWR

    Energy Technology Data Exchange (ETDEWEB)

    Tavera D, L; Camacho L, M E

    1991-01-15

    The measurements carried out on reactor dosimetry are applied mainly to the study on the effects of the radiation in 108 materials of the reactor; little is on the environmental dosimetry outside of the primary container of BWR reactors. In this work the application of a neutron spectrometer formed by plastic detectors of nuclear traces manufactured in the ININ, for the environmental monitoring in penetrations around the primary container of the unit I of the Laguna Verde central is presented. The neutron monitoring carries out with purposes of radiological protection, during the operational tests of the reactor. (Author)

  1. 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

  2. Twenty-fifth water reactor safety information meeting: Proceedings. Volume 1: Plenary sessions; Pressure vessel research; BWR strainer blockage and other generic safety issues; Environmentally assisted degradation of LWR components; Update on severe accident code improvements and applications

    International Nuclear Information System (INIS)

    Monteleone, S.

    1998-03-01

    This three-volume report contains papers presented at the conference. 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 France, Japan, Norway, and Russia. The titles of the papers and the names of the authors have been updated and may differ from those that appeared in the final program of the meeting. This volume contains the following information: (1) plenary sessions; (2) pressure vessel research; (3) BWR strainer blockage and other generic safety issues; (4) environmentally assisted degradation of LWR components; and (5) update on severe accident code improvements and applications. Selected papers have been indexed separately for inclusion in the Energy Science and Technology Database

  3. Twenty-fifth water reactor safety information meeting: Proceedings. Volume 1: Plenary sessions; Pressure vessel research; BWR strainer blockage and other generic safety issues; Environmentally assisted degradation of LWR components; Update on severe accident code improvements and applications

    Energy Technology Data Exchange (ETDEWEB)

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

    1998-03-01

    This three-volume report contains papers presented at the conference. 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 France, Japan, Norway, and Russia. The titles of the papers and the names of the authors have been updated and may differ from those that appeared in the final program of the meeting. This volume contains the following information: (1) plenary sessions; (2) pressure vessel research; (3) BWR strainer blockage and other generic safety issues; (4) environmentally assisted degradation of LWR components; and (5) update on severe accident code improvements and applications. Selected papers have been indexed separately for inclusion in the Energy Science and Technology Database.

  4. 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

  5. Reactor pressure vessel status report

    International Nuclear Information System (INIS)

    Strosnider, J.; Wichman, K.; Elliot, B.

    1994-12-01

    This report gives a brief description of the reactor pressure vessel (RPV), followed by a discussion of the radiation embrittlement of RPV beltline materials and the two indicators for measuring embrittlement, the end-of-license (EOL) reference temperature and the EOL upper-shelf energy. It also summarizes the GL 92-01 effort and presents, for all 37 boiling water reactor plants and 74 pressurized water reactor plants in the United States, the current status of compliance with regulatory requirements related to ensuring RPV integrity. The staff has evaluated the material data needed to predict neutron embrittlement of the reactor vessel beltline materials. These data will be stored in a computer database entitled the reactor vessel integrity database (RVID). This database will be updated annually to reflect the changes made by the licensees in future submittals and will be used by the NRC staff to assess the issues related to vessel structural integrity

  6. Residual stress analysis in BWR pressure vessel attachments

    International Nuclear Information System (INIS)

    Dexter, R.J.; Leung, C.P.; Pont, D.

    1992-06-01

    Residual stresses from welding processes can be the primary driving force for stress corrosion cracking (SCC) in BWR components. Thus, a better understanding of the causes and nature of these residual stresses can help assess and remedy SCC. Numerical welding simulation software, such as SYSWELD, and material property data have been used to quantify residual stresses for application to SCC assessments in BWR components. Furthermore, parametric studies using SYSWELD have revealed which variables significantly affect predicted residual stress. Overall, numerical modeling techniques can be used to evaluate residual stress for SCC assessments of BWR components and to identify and plan future SCC research

  7. Flow rate control systems for coolants for BWR type reactors

    International Nuclear Information System (INIS)

    Igarashi, Yoko; Kato, Naoyoshi.

    1981-01-01

    Purpose: To increase spontaneous recycling flow rate of coolants in BWR type reactors when the water level in the reactor decreases, by communicating a downcomer with a lower plenum. Constitution: An opening is provided to the back plate disposed at the lower end of a reactor core shroud for communicating a downcomer with a lower plenum, and an ON-OFF valve actuated by an operation rod is provided to the opening. When abnormal water level or pressure in the reactor is detected by a level metal or pressure meter, the operation rod is driven to open the ON-OFF valve, whereby coolants fed from a jet pump partially flows through the opening to increase the spontaneous recycling flow rate of the coolants. This can increase the spontaneous recycling flow rate of the coolants upon spontaneous recycling operation, thereby maintaining the reactor safety and the fuel soundness. (Moriyama, K.)

  8. Radiation streaming in power reactors. [PWR; BWR

    Energy Technology Data Exchange (ETDEWEB)

    Lahti, G.P.; Lee, R.R.; Courtney, J.C. (eds.)

    1979-02-01

    Separate abstracts are included for each of the 14 papers given at a special session on Radiation Streaming in Power Reactors held on November 15 at the American Nuclear Society 1978 Winter Meeting in Washington, D.C. The papers describe the methods of calculation, the engineering of shields, and the measurement of radiation environments within the containments of light water power reactors. Comparisons of measured and calculated data are used to determine the accuracy of computer predictions of the radiation environment. Specific computational and measurement techniques are described and evaluated. Emphasis is on radiation streaming in the annular region between the reactor vesel and the primary shield and its resultant environment within the primary containment.

  9. Power controlling method for BWR type reactors

    International Nuclear Information System (INIS)

    Yoshida, Kenji.

    1983-01-01

    Purpose: To enable reactor operation exactly following after an aimed curve in the high power resuming and maintaining period without failures in cladding tubes. Method: Upon recovery of the reactor power to a high power level after changing the reactor power from the high power to the low power level, control rod is operated under such conditions that the linear power density after operation of the control rod does not exceed the PC envelope in the low power period, and the core flow rate is coordinated to the control rod operation. The linear power density can be suppressed within an allowable linear power density by the above operation during high power resuming and maintaining period and, as the result, PCI failures can be prevented. (Kamimura, M.)

  10. Reactor-vessel-sectioning demonstration

    International Nuclear Information System (INIS)

    Lundgren, R.A.

    1981-07-01

    A successful technical demonstration of simulated reactor vessel sectioning was completed 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: an air arc gouger was chosen to penetrate the stainless steel, and flame cutting was selected to sever the carbon steel. After the simulated vessel was successfully cut from the SS side, another cut was made, starting from the carbon steel side. This cut was also successful. Cutting from the carbon steel side has the advantages of cost reduction since the air arc gouging step is eliminated and contamination controlled because the molten metal is blown inward

  11. Analytical and experimental vibration analysis of BWR pressure vessel internals

    International Nuclear Information System (INIS)

    Krutzik, N.; Schad, O.

    1975-01-01

    This report attempts to evaluate the validity as well as quality of several analytical methods in the light of presently available experimental data for the internals of pressure vessels of boiling-water-reactor-types. The experimental checks were performed after the numerical analysis was completed and showed the accuracy of the numerical results. The analytical investigations were done by finite element programmes - 2-dimensional as well as 3-dimensional, where the effect of the mass distribution with parts of virtual masses on the dynamic response could be studied in depth. The experimental data were collected at various different plants and with different mass correlations. Besides evaluating the dynamic characteristics of the components, tests were also performed to evaluate the vibrations of the pressure vessel relative to the main structure. After analysing extensive recorded data much better understanding of the response under a variety of loading- and boundary conditions could be gained. The comparison of the results of analytical studies with the experimental results made a broad qualitative evaluation possible. (Auth.)

  12. Operation control equipment for BWR type reactor

    International Nuclear Information System (INIS)

    Izumi, Masayuki; Takeda, Renzo.

    1981-01-01

    Purpose: To improve the temperature balance in a feedwater heater by obtaining the objective value of a feedwater enthalpy upon calculation of respective measured values and controlling the opening or closing of an extraction valve so that the objective value may coincide with the measured value, thereby averaging the axial power distribution. Constitution: A plurality of stages of extraction lines are connected to a turbine, and extraction valves are respectively provided at the lines. By calculating the measured values of ractor pressure, reactor core flow rate, vapor flow rate and reactor core inlet enthalpy determined to predetermined value using heat balance the objective feedwater enthalpy is obtained, is fed as an extraction valve opening or closing signal from a control equipment, the extraction stages of the turbine extraction are altered in accordance with this signal, and the feedwater enthalpy is controlled. (Sekiya, K.)

  13. Core arrangement in BWR type reactors

    International Nuclear Information System (INIS)

    Asano, Masayuki.

    1981-01-01

    Purpose: To decrease the number of fuel assemblies whose locations are to be changed upon fuel exchange, as well as unify the power distribution in the core by arranging, in a chess board configuration, a plurality pattern of unit reactor lattices each containing fuel assemblies of different burnup degrees in orthogonal positions to each other. Constitution: A first pattern of unit reactor lattice is formed by disposing fuel assemblies of burnup degree 1 and fuel assemblies of burnup degree 3 at orthogonal positions to each other. A second pattern of unit reactor lattice is formed by disposing fuel assemblies of burnup degree 2 and fuel assemblies of burnup degree 1 at orthogonal positions to each other. The unit lattices each in such a dispositions are arranged in a chess board arrangement. Since, the fuel assemblies of the burnup degree 1 in the first pattern unit lattices proceed to the burnup degree 2 and the fuel assemblies of the burnup degree 2 in the second pattern unit lattices proceed to the burnup degree 3 up to the fuel exchange stage, fuel exchange and movement have only to be made, not for those fuel assemblies, but for another half of the fuel assemblies. (Kawakami, Y.)

  14. Primary cooling system for BWR type reactor

    International Nuclear Information System (INIS)

    Ibe, Eishi; Takahashi, Masanori; Aoki, Yasuko

    1993-01-01

    The present invention effectively uses information from a plurality of sensors in order to suppress corrosion circumstance of a nuclear reactor. That is, a predetermined general water quality factor at a predetermined position is determined as a standard index. A concentration of a water quality improver is controlled such that the index is within an aimed range. For this purpose, the entire sensor groups disposed in a primary coolant system of a nuclear reactor are divided into a plural systems of sensor groups each disposed on every different positions. Then, a predetermined sensor group (standard sensor group) is connected to a computing device and a data base so that it is always monitored for calculating and estimating the standard index. Only oxidative ingredient in water at the measuring point is noted, and a concentration distribution which agrees with an actually measured value of oxidative ingredients is extracted from data base and used as a correct concentration distribution. With such procedures, reactor water quality can be estimated accurately while compensating erroneous factors of individual sensors. Even when a new sensor is used, it is not necessary to greatly change control logic. (I.S.)

  15. 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

  16. BWR type reactor and its operating method

    International Nuclear Information System (INIS)

    Ootsuji, Niro.

    1983-01-01

    Purpose: To regulate the control rod extraction operation such that an assumed control rod drop accident, if should occur, may not lead to further serious accidents, as well as enable to improve the working life of the control rod. Method: A plurality of control rods disposed among a plurality of fuel assemblies constituting the reactor core for suppressing the reactor core reactivity are divided into two groups depending on the descending speed, and the number of rods with a faster descending speed is set to less than 1/4 of the total number of the control rods. Then, the control rods are arranged such that those rods of the faster descending speed may be set every one another in any of the vertical, lateral and orthogonal directions. Further, it is always judged as to the possibility of extracting the control rods with the faster descending speed by a fast control rod extraction judging circuit to issue a signal to a control rod extraction inhibition circuit, so that the extraction operation for the control rods with the faster descending speed is started after all of the control rods with the slow descending speed have been extracted. Accordingly, if a control rod dropping accident should occur, abrupt power change can be avoided to thereby minimize the development of the accident. (Horiuchi, T.)

  17. Corrosion failure of a BWR embedded reactor containment liner

    International Nuclear Information System (INIS)

    Wegemar, B.

    2006-01-01

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

  18. 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.

  19. 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)

  20. Latest experiences in inspecting the inside of BWR vessel shields

    Energy Technology Data Exchange (ETDEWEB)

    Alberdi, R.; Gonzalez, E.

    2001-07-01

    In the last few years, the owners of BWR nuclear power plants have been forced to address new fuel shield inspection requirements, TECNATOM has responded to this situation by launching the TEIDE projects, which include development of an inspection machine and the corresponding Non-Destructive Tests to examine the inside of this shield. With these projects, TECNATOM has performed more than 12 fuel shield inspections in different countries. This article describes the experience gained in the last three years. (Author)

  1. Fuel assembly for BWR type reactor

    International Nuclear Information System (INIS)

    Kato, Shigeru.

    1993-01-01

    In the fuel assembly of the present invention, a means for mounting and securing short fuel rods is improved. Not only long fuel rods but also short fuel rods are disposed in channel of the fuel assembly to improve reactor safety. The short fuel rods are supported by a screw means only at the lower end plug. The present invention prevents the support for the short fuel rod from being unreliable due to the slack of the screw by the pressure of inflowing coolants. That is, coolant abutting portions such as protrusions or concave grooves are disposed at a portion in the channel box where coolants flowing from the lower tie plate, as an uprising stream, cause collision. With such a constitution, a component caused by the pressure of the flowing coolants is formed. The component acts as a rotational moment in the direction of screwing the male threads of the short fuel rod into the end plug screw hole. Accordingly, the screw is not slackened, and the short fuel rods are mounted and secured certainly. (I.S.)

  2. 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

  3. Coolant make-up device for BWR type reactor

    International Nuclear Information System (INIS)

    Sasagawa, Hiroshi.

    1994-01-01

    In a coolant make-up device, an opening of a pressure equalizing pipeline in a pressure vessel is disposed in coolants above a reactor core and below a usual fluctuation range of a reactor vessel water level. Further, a float check valve is disposed to the pressure equalizing pipeline for preventing coolants in the pressure vessel flowing into the pipeline. If the water level in the pressure vessel is lowered than the setting position for the float check valve, the float drops by its own weight to open the opening of the pressure equalizing pipeline. Then, steams in the pressure vessel are flown into the pipeline, to equalize the pressure between a coolant storage tank and the pressure vessel of the reactor. Coolants in the coolant storage tank is injected to the pressure vessel by way of the water injection pipeline due to the difference of the pressure head between the water level in the coolants storage tank and the water level in the pressure vessel. If the coolants are lowered than the setting position for the float check value, the float check valve does not close unless the water level is recovered to the setting position for the float valve and, accordingly, the coolant make-up is continued. (N.H.)

  4. Simulation of the automatic depressurization system (Ads) for a boiling water reactor (BWR) based on RELAP

    International Nuclear Information System (INIS)

    Ramirez G, C.; Chavez M, C.

    2012-10-01

    The automatic depressurization system (Ads) of the boiling water reactor (BWR) like part of the emergency cooling systems is designed to liberate the vapor pressure of the reactor vessel, as well as the main vapor lines. At the present time in the Engineering Faculty, UNAM personnel works in the simulation of the Laguna Verde reactor based on the nuclear code RELAP/SCADAP and in the incorporation to the same of the emergency cooling systems. The simulation of the emergency cooling systems began with the inclusion of two hydrodynamic volumes, one source and another drain, and the incorporation of the initiation logic for each emergency system. In this work is defined and designed a simplified model of Ads of the reactor, considering a detail level based on the main elements that compose it. As tool to implement the proposed model, the RELAP code was used. The simulated main functions of Ads are centered in the quick depressurization of the reactor by means of the vapor discharge through the relief/safety valves to the suppression pool, and, in the event of break of the main vapor line, the reduction of the vessel pressure operates for that the cooling systems of the core to low pressure (Lpcs and Lpci) they can begin their operation. (Author)

  5. Generic aging management programs for license renewal of BWR reactor coolant systems components

    International Nuclear Information System (INIS)

    Shah, V.N.; Liu, Y.Y.

    2002-01-01

    The paper reviews the existing generic aging management programs (AMPs) for the reactor coolant system (RCS) components in boiling water reactors (BWRs), including the reactor pressure vessel and internals, the reactor recirculation system, and the connected piping. These programs have been evaluated in the U.S. Nuclear Regulatory Commission (NRC) report, Generic Aging Lessons Learned (GALL), NUREG-1801, for their use in the license renewal process to manage several aging effects, including loss of material, crack initiation and growth, loss of fracture toughness, loss of preload, wall thinning, and cumulative fatigue damage. The program evaluation includes a review of ten attributes (scope of program, preventive actions, parameters monitored/inspected, detection of aging effects, monitoring and trending, acceptance criteria, corrective actions, confirmative process, administrative control, and operating experience) for their effectiveness in managing a specific aging effect in a given component(s). The generic programs are based on the ASME Section XI inservice inspection requirements; industry guidelines for inspection and evaluation of aging effects in BWR reactor vessel, internals, and recirculation piping; monitoring and control of BWR water chemistry; and operating experience as reported in the USNRC generic communications and industry reports. The review concludes that all generic AMPs are acceptable for managing aging effects in BWR RCS components during an extended period of operation and do not need further evaluation. However, the plant-specific programs for managing aging in certain RCS components during an extended period of operation do require further evaluation. For some plant-specific AMPs, the GALL report recommends an aging management activity to verify their effectiveness. An example of such an activity is a one-time inspection of Class 1 small-bore piping to ensure that service-induced weld cracking is not occurring in the piping. Several of

  6. Generic Aging Management Programs for License Renewal of BWR Reactor Coolant System Components

    International Nuclear Information System (INIS)

    Shah, V.N.; Liu, Y.Y.

    2002-01-01

    The paper reviews the existing generic aging management programs (AMPs) for the reactor coolant system (RCS) components in boiling water reactors (BWRs), including the reactor pressure vessel and internals, the reactor recirculation system, and the connected piping. These programs have been evaluated in the U.S. Nuclear Regulatory Commission (NRC) report, Generic Aging Lessons Learned (GALL), NUREG-1801, for their use in the license renewal process to manage several aging effects, including loss of material, crack initiation and growth, loss of fracture toughness, loss of preload, wall thinning, and cumulative fatigue damage. The program evaluation includes a review of ten attributes (scope of program, preventive actions, parameters monitored/inspected, detection of aging effects, monitoring and trending, acceptance criteria, corrective actions, confirmative process, administrative control, and operating experience) for their effectiveness in managing a specific aging effect in a given component(s). The generic programs are based on the ASME Section XI inservice inspection requirements; industry guidelines for inspection and evaluation of aging effects in BWR reactor vessel, internals, and recirculation piping; monitoring and control of BWR water chemistry; and operating experience as reported in the USNRC generic communications and industry reports. The review concludes that all generic AMPs are acceptable for managing aging effects in BWR RCS components during an extended period of operation and do not need further evaluation. However, the plant-specific programs for managing aging in certain RCS components during an extended period of operation do require further evaluation. For some plant-specific AMPs, the GALL report recommends an aging management activity to verify their effectiveness. An example of such an activity is a one-time inspection of Class 1 small-bore piping to ensure that service-induced weld cracking is not occurring in the piping. Several of

  7. Artificial intelligence applied to fuel management in BWR type reactors

    International Nuclear Information System (INIS)

    Ortiz S, J.J.

    1998-01-01

    In this work two techniques of artificial intelligence, neural networks and genetic algorithms were applied to a practical problem of nuclear fuel management; the determination of the optimal fuel reload for a BWR type reactor. This is an important problem in the design of the operation cycle of the reactor. As a result of the application of these techniques, comparable or even better reloads proposals than those given by expert companies in the subject were obtained. Additionally, two other simpler problems in reactor physics were solved: the determination of the axial power profile and the prediction of the value of some variables of interest at the end of the operation cycle of the reactor. Neural networks and genetic algorithms have been applied to solve many problems of engineering because of their versatility but they have been rarely used in the area of fuel management. The results obtained in this thesis indicates the convenience of undertaking further work on this area and suggest the application of these techniques of artificial intelligence to the solution of other problems in nuclear reactor physics. (Author)

  8. Performance of evaluation techniques for planning preventive repair of leakages on BWR vessel stub tubes

    International Nuclear Information System (INIS)

    Prat, J.P.

    1985-01-01

    Stub tubes have been installed on the control rod drive assemblies of the majority of boiling water reactor vessel bottoms. These stub tubes can undergo a slow process of cracking, producing no potential hazard but occasional plant unavailability due to leakage in excess of specification limits. In the case considered here, the development and application of remotely controlled equipment permitting access both from inside and outside the housing, and of crack evolution and location techniques, has made the progressive development, selection and planning of repairs a possibility, thus minimizing the risk of leakages appearing between refuelling outages. The example used here is of a Nuclear Power Plant in which a first leak appeared in 1981. Since that time extensive and progressive laboratory and field development work has been carried out. This paper presents the experience gained with customized, remote-control inspection equipment, and several developed or adapted techniques, as the basis for an in-service repair decision tree, the case in question being that of a BWR plant in which a cracking process, not affecting safety but potentially affecting availability, has given rise to leaks after ten years of operation

  9. 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

  10. Calculation device for fuel power history in BWR type reactors

    International Nuclear Information System (INIS)

    Sakagami, Masaharu.

    1980-01-01

    Purpose: To enable calculations for power history and various variants of power change in the power history of fuels in a BWR type reactor or the like. Constitution: The outputs of the process computation for the nuclear reactor by a process computer are stored and the reactor core power distribution is judged from the calculated values for the reactor core power distribution based on the stored data. Data such as for thermal power, core flow rate, control rod position and power distribution are recorded where the changes in the power distribution exceed a predetermined amount, and data such as for thermal power and core flow rate are recorded where the changes are within the level of the predetermined amount, as effective data excluding unnecessary data. Accordingly, the recorded data are taken out as required and the fuel power history and the various variants in the fuel power are calculated and determined in a calculation device for fuel power history and variants for fuel power fluctuation. (Furukawa, Y.)

  11. Identification of the reduced order models of a BWR reactor

    International Nuclear Information System (INIS)

    Hernandez S, A.

    2004-01-01

    The present work has as objective to analyze the relative stability of a BWR type reactor. It is analyzed that so adaptive it turns out to identify the parameters of a model of reduced order so that this it reproduces a condition of given uncertainty. This will take of a real fact happened in the La Salle plant under certain operation conditions of power and flow of coolant. The parametric identification is carried out by means of an algorithm of recursive least square and an Output Error model (Output Error), measuring the output power of the reactor when the instability is present, and considering that it is produced by a change in the reactivity of the system in the same way that a sign of type step. Also it is carried out an analytic comparison of the relative stability, analyzing two types of answers: the original answer of the uncertainty of the reactor vs. the obtained response identifying the parameters of the model of reduced order, reaching the conclusion that it is very viable to adapt a model of reduced order to study the stability of a reactor, under the only condition to consider that the dynamics of the reactivity is of step type. (Author)

  12. Nuclear reactor construction with bottom supported reactor vessel

    International Nuclear Information System (INIS)

    Sharbaugh, J.E.

    1987-01-01

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

  13. Emergency core cooling system in BWR type reactors

    International Nuclear Information System (INIS)

    Takizawa, Yoji

    1981-01-01

    Purpose: To rapidly recover the water level in the reactor upon occurrence of slight leakages in the reactor coolant pressure boundary, by promoting the depressurization in the reactor to thereby rapidly increase the high pressure core spray flow rate. Constitution: Upon occurrence of reactor water level reduction, a reactor isolation cooling system and a high pressure core spray system are actuated to start the injection of coolants into a reactor pressure vessel. In this case, if the isolation cooling system is failed to decrease the flow rate in a return pipeway, flow rate indicators show a lower value as compared with a predetermined value. The control device detects it and further confirms the rotation of a high pressure spray pump to open a valve. By the above operation, coolants pumped by the high pressure spray pump is flown by way of a communication pipeway to the return pipeway and sprayed from the top of the pressure vessel. This allows the vapors on the water surface in the pressure vessel to be cooled rapidly and increases the depressurization effects. (Horiuchi, T.)

  14. 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

  15. Compact modular BWR (CM-BWR)

    International Nuclear Information System (INIS)

    Fennern, Larry; Boardman, Charles; Carroll, Douglas G.; Hida, Takahiko

    2003-01-01

    A preliminary assessment has shown that a small 350 MWe BWR reactor can be placed within a close fitting steel containment vessel that is 7.1 meters inside diameter. This allows the technology and manufacturing capability currently used to fabricate large ABWR reactor vessels to be used to provide a factory fabricated containment vessel for a 350 MWe BWR. When a close fitted steel containment is combined with a passive closed loop isolation condenser system and a natural circulating reactor system that contains a large water inventory, primary system leaks cannot uncover the core. This eliminates many of the safety systems needed in response to a LOCA that are common to large, conventional plant designs including. Emergency Core Flooding, Automatic Depressurization System, Active Residual Heat Removal, Safety Related Auxiliary Cooling, Safety Related Diesel Generators, Hydrogen Re-Combiners, Ex-vessel Core Retention and Cooling. By fabricating the containment in a factory and eliminating most of the conventional safety systems, the construction schedule is shortened and the capital cost reduced to levels that would not otherwise be possible for a relatively small modular BWR. This makes the CM-BWR a candidate for applications where smaller incremental power additions are desired relative to a large ALWR or where the local infrastructure is not able to accommodate a conventional ALWR plant rated at 1350 MWe or more. This paper presents a preliminary design description of a Compact Modular BWR (CM-BWR) whose design features dramatically reduce the size and cost of the reactor building and associated safety systems. (author)

  16. Identification of the reduced order models of a BWR reactor; Identificacion de modelos de orden reducido de un reactor BWR

    Energy Technology Data Exchange (ETDEWEB)

    Hernandez S, A. [UNAM, Laboratorio de Analisis de Ingenieria de Reactores Nucleares, DEPFI, Campus Morelos, en IMTA Jiutepec, Morelos (Mexico)]. e-mail: augusto@correo.unam.mx

    2004-07-01

    The present work has as objective to analyze the relative stability of a BWR type reactor. It is analyzed that so adaptive it turns out to identify the parameters of a model of reduced order so that this it reproduces a condition of given uncertainty. This will take of a real fact happened in the La Salle plant under certain operation conditions of power and flow of coolant. The parametric identification is carried out by means of an algorithm of recursive least square and an Output Error model (Output Error), measuring the output power of the reactor when the instability is present, and considering that it is produced by a change in the reactivity of the system in the same way that a sign of type step. Also it is carried out an analytic comparison of the relative stability, analyzing two types of answers: the original answer of the uncertainty of the reactor vs. the obtained response identifying the parameters of the model of reduced order, reaching the conclusion that it is very viable to adapt a model of reduced order to study the stability of a reactor, under the only condition to consider that the dynamics of the reactivity is of step type. (Author)

  17. The World's Reactors no. 70 - Forsmark 3, BWR-75

    International Nuclear Information System (INIS)

    Anon.

    1976-01-01

    A large pull-out wall chart is presented showing a coloured cut-away diagram of the Forsmark 3 station. It is accompanied by 2 small sketches one showing the layout of station buildings and the other the inside of the reactor vessel. Parameters are listed. (U.K.)

  18. Studies on core melt behaviour in a BWR pressure vessel lower head

    International Nuclear Information System (INIS)

    Lindholm, I.; Ikonen, K.; Hedberg, K.

    1999-01-01

    Core debris behaviour in the Nordic BWR lower head was investigated numerically using MELCOR and MAAP4 codes. Lower head failure due to penetration failure was studied with more detailed PASULA code taking thermal boundary conditions from MELCOR calculations. Creep rupture failure mode was examined with the two integral codes. Also, the possibility to prevent vessel failure by late reflooding was assessed in this study. (authors)

  19. Investigation of decreasing reactor coolant inventory as a mechanism to reduce power during a BWR ATWS

    International Nuclear Information System (INIS)

    Peterson, C.E.; Chexal, V.K.; Layman, W.; Hentzen, R.D.; Gose, G.C.

    1985-01-01

    A best-estimate analysis was performed to evaluate the technique of intentionally reducing reactor coolant inventory in order to reduce power during a BWR ATWS event. The ATWS was initiated by closure of the main steam isolation valves. The analysis was performed with the RETRAN-02 computer code utilizing the one-dimensional kinetics option. The one-dimensional cross sections were developed using the SIMULATE-E and SIMTRAN-E computer codes. The MSIV closure transient provides some of the more severe conditions following a postulated failure to scram. In this transient, the only mechanism for removing energy from the vessel is through the safety relief valve system which results in a heating up of the suppression pool fluid. Consequently, the reactor power must be reduced so that the suppression pool temperature limits are not exceeded. Under the proposed emergency procedure guidelines for the ATWS event, the reactor vessel water level will be lowered to reduce system power. This analysis evaluated the dynamic response of the system as the water level was lowered to the top of active fuel evaluation. Correlating the system power and flow patterns to water level was of particular interest in the analysis. Under natural circulating conditions, the system flows, core power, and pressure responses are extremely tightly coupled and the analysis results proved to be very sensitive to the modeling of downcomer, upper plenum, and core regions

  20. 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

  1. Reactor power control device in BWR power plant

    International Nuclear Information System (INIS)

    Kurosawa, Tsuneo.

    1997-01-01

    The present invention provides a device for controlling reactor power based on a start-up/shut down program in a BWR type reactor, as well as for detecting deviation, if occurs, of the power from the start-up/shut down program, to control a recycling flow rate control system or control rod drive mechanisms. Namely, a power instruction section successively executes the start-up/shut down program and controls the coolant recycling system and the control rod driving mechanisms to control the power. A current state monitoring and calculation section receives a process amount, calculates parameters showing the plant state, compares/monitors them with predetermined values, detecting the deviation, if occurs, of the plant state from the start-up/shut down program, and prevents output of a power increase control signal which leads to power increase. A forecasting and monitoring/calculation section forecasts and calculates the plant state when not yet executed steps of the start-up/shut down program are performed, stops the execution of the start-up/shut down program in the next step in a case of forecasting that the results of the calculation will deviate from the start-up/shut down program. (I.S.)

  2. Design of a redundant meteorological station for a BWR reactor

    International Nuclear Information System (INIS)

    Ramirez S, R.; Celis del Angel, L.; Bucio, F.; Rivero, T.; Palacios, J.

    2008-01-01

    In this work the design of a meteorological station for a reactor type BWR is proposed. Two independent channels of data acquisition that allow him to have a bigger readiness is exposed. It is incorporate sensors without mobile parts to measure speed, wind direction and pluvial precipitation. It also counts, with sensors of global solar radiation, net radiation, barometric pressure, relative humidity and ambient temperature; with them they are possible to be calculated, moreover, other variables as temperature differential, dew point and atmospheric stability. The sensors are placed on a tower to different heights and send their information (each second) to a local registration system, the one which in turn, it remits the data to the monitoring office so that a computer is linked with the system, display and management the information in real time and automatic way. The redundant structure allows that in the event of maintenance the data acquisition is not interrupted, even if the information is transferred to another place. In all the station sections it is used protocols of standard communication to allow that a great quantity of devices can be connected without major problem. The above-mentioned would allow to the operators in the control room to have reliable information during the whole time of the reactor operation. (Author)

  3. Hybrid Reactor Simulation and 3-D Information Display of BWR Out-of-Phase Oscillation

    International Nuclear Information System (INIS)

    Edwards, Robert; Huang, Zhengyu

    2001-01-01

    The real-time hybrid reactor simulation (HRS) capability of the Penn State TRIGA reactor has been expanded for boiling water reactor (BWR) out-of-phase behavior. During BWR out-of-phase oscillation half of the core can significantly oscillate out of phase with the other half, while the average power reported by the neutronic instrumentation may show a much lower amplitude for the oscillations. A description of the new HRS is given; three computers are employed to handle all the computations required, including real-time data processing and graph generation. BWR out-of-phase oscillation was successfully simulated. By adjusting the reactivity feedback gains from boiling channels to the TRIGA reactor and to the first harmonic mode power simulation, limit cycle can be generated with both reactor power and the simulated first harmonic power. A 3-D display of spatial power distributions of fundamental mode, first harmonic, and total powers over the reactor cross section is shown

  4. 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

  5. Improved nuclear reactor construction with bottom supported reactor vessel

    International Nuclear Information System (INIS)

    Sharbaugh, J.E.

    1987-01-01

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

  6. 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

  7. 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.

  8. Development of the radiation models of a BWR type reactor and it facility in the SUN-RAH; Desarrollo de modelos de radiacion de un reactor tipo BWR y su instalacion en el SUN-RAH

    Energy Technology Data Exchange (ETDEWEB)

    Barron A, I. [Facultad de Ingenieria, UNAM, 04510 Mexico D.F. (Mexico)]. e-mail: isbarron@yahoo.com.mx

    2005-07-01

    This work about generation models, transport in processes and radioactive contamination of areas of a BWR central, is an amplification to the project developed in the UNAM to have a support tool in subjects or electric generation courses. It is planned about the implementation of models of radiation generation in a BWR type reactor for complement the functions developed in the University Simulator of Nucleo electric- Boiling water reactor (SUN-RAH) which it has been implemented in Simulink of MatLab and it has a model for the dynamics of one nucleo electric central that presents the main characteristics of the reactor vessel, the recirculation system, steam lines, turbines, generator, condensers and feeding water, defined by the main processes that intervene in the generation of energy of these plants. By this way the radiation monitoring systems for area and process, operate simultaneously with the processes of energy generation, with that is possible to observe the changes that present with respect to the operation conditions of the plant, and likewise to appreciate the radiation transport process through the components of the reactor, steam lines and turbines, for different operation conditions and possible faults that they could be presented during the reactor operation. (Author)

  9. Repairing method for shroud in reactor pressure vessel

    International Nuclear Information System (INIS)

    Watanabe, Yusuke.

    1996-01-01

    The present invention provides a method of repairing a shroud disposed in a pressure vessel of a BWR type reactor. Namely, a baffle plate is disposed on the outer surface of the lower portion of the shroud supported by a shroud support of the pressure vessel. The baffle plate is connected with a lug for securing a shroud head bolt disposed on the outer surface of an upper portion of the shroud by reinforcing members. With such a constitution, when crackings are caused in the shroud, the development of the crackings can be prevented without losing the function of securing the shroud head bolt. Further, if a material having thermal expansion coefficient lower than that of austenite stainless steel is used for the material of the reinforcing member, clamping load to be applied upon attaching the auxiliary member can be reduced. As a result, operation for the attachment is facilitated. (I.S.)

  10. Fuel assemblies for use in BWR type reactors

    International Nuclear Information System (INIS)

    Hirukawa, Koji.

    1987-01-01

    Purpose: To moderate the peak configuration of the burnup degree change curve for the infinite multiplication factor by applying an improvement to the arrangement of fuel rods. Constitution: In a fuel assembly for a BWR type reactor comprising a plurality of fuel rods and water rods arranged in a square lattice, fuel rods containing burnable poisons are arranged at four corners at the second and the third layers from the outside of the square lattice arrangement. Among them, the Cd poison effect in the burnable poison incorporated fuel rods disposed at the second layer is somewhat greater at the initial burning stage and then rapidly decreased along with burning. While on the other hand, the poison effect of the burnable poison-incorporated fuel rods at the third layer is smaller than that at the second layer at the initial burning stage and the reduction in the poison effect due to burning is somewhat more moderate. Since these fuel rods are in adjacent with each other, they interfere to each other and also provide an effect of moderating the burning of the burnable poisons. (Takahashi, M.)

  11. Development of the radiation models of a BWR type reactor and it facility in the SUN-RAH

    International Nuclear Information System (INIS)

    Barron A, I.

    2005-01-01

    This work about generation models, transport in processes and radioactive contamination of areas of a BWR central, is an amplification to the project developed in the UNAM to have a support tool in subjects or electric generation courses. It is planned about the implementation of models of radiation generation in a BWR type reactor for complement the functions developed in the University Simulator of Nucleo electric- Boiling water reactor (SUN-RAH) which it has been implemented in Simulink of MatLab and it has a model for the dynamics of one nucleo electric central that presents the main characteristics of the reactor vessel, the recirculation system, steam lines, turbines, generator, condensers and feeding water, defined by the main processes that intervene in the generation of energy of these plants. By this way the radiation monitoring systems for area and process, operate simultaneously with the processes of energy generation, with that is possible to observe the changes that present with respect to the operation conditions of the plant, and likewise to appreciate the radiation transport process through the components of the reactor, steam lines and turbines, for different operation conditions and possible faults that they could be presented during the reactor operation. (Author)

  12. 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)

  13. 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.)

  14. 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

  15. 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

  16. Design and axial optimization of nuclear fuel for BWR reactors

    International Nuclear Information System (INIS)

    Garcia V, M.A.

    2006-01-01

    In the present thesis, the modifications made to the axial optimization system based on Tabu Search (BT) for the axial design of BWR fuel type are presented, developed previously in the Nuclear Engineering Group of the UNAM Engineering Faculty. With the modifications what is mainly looked is to consider the particular characteristics of the mechanical design of the GE12 fuel type, used at the moment in the Laguna Verde Nucleo electric Central (CNLV) and that it considers the fuel bars of partial longitude. The information obtained in this thesis will allow to plan nuclear fuel reloads with the best conditions to operate in a certain cycle guaranteeing a better yield and use in the fuel burnt, additionally people in charge in the reload planning will be favored with the changes carried out to the system for the design and axial optimization of nuclear fuel, which facilitate their handling and it reduces their execution time. This thesis this developed in five chapters that are understood in the following way in general: Chapter 1: It approaches the basic concepts of the nuclear energy, it describes the physical and chemical composition of the atoms as well as that of the uranium isotopes, the handling of the uranium isotope by means of the nuclear fission until arriving to the operation of the nuclear reactors. Chapter 2: The nuclear fuel cycle is described, the methods for its extraction, its conversion and its enrichment to arrive to the stages of the nuclear fuel management used in the reactors are described. Beginning by the radial design, the axial design and the core design of the nuclear reactor related with the fuel assemblies design. Chapter 3: the optimization methods of nuclear fuel previously used are exposed among those that are: the genetic algorithms method, the search methods based on heuristic rules and the application of the tabu search method, which was used for the development of this thesis. Chapter 4: In this part the used methodology to the

  17. Prevention device for rapid reactor core shutdown in BWR type reactors

    International Nuclear Information System (INIS)

    Koshi, Yuji; Karatsu, Hiroyuki.

    1986-01-01

    Purpose: To surely prevent rapid shutdown of a nuclear reactor upon partial load interruption due to rapid increase in the system frequency. Constitution: If a partial load interruption greater than the sum of the turbine by-pass valve capacity and the load setting bias portion is applied in a BWR type power plant, the amount of main steams issued from the reactor is decreased, the thermal input/output balance of the reactor is lost, the reactor pressure is increased, the void is collapsed, the neutron fluxes are increased and the reactor power rises to generate rapid reactor shutdown. In view of the above, the turbine speed signal is compared with a speed setting value in a recycling flowrate control device and the recycling pump is controlled to decrease the recycling flowrate in order to compensate the increase in the neutron fluxes accompanying the reactor power up. In this way, transient changes in the reactor core pressure and the neutron fluxes are kept within a setting point for the rapid reactor shutdown operation thereby enabling to continue the plant operation. (Horiuchi, T.)

  18. Applied methods for mitigation of damage by stress corrosion in BWR type reactors; Metodos aplicados para la mitigacion del dano por corrosion bajo esfuerzo en reactores BWR

    Energy Technology Data Exchange (ETDEWEB)

    Hernandez C, R.; Diaz S, A.; Gachuz M, M.; Arganis J, C. [Instituto Nacional de Investigaciones Nucleares, Gerencia de Ciencia de Materiales, A.P. 18-1027, 11801 Mexico D.F. (Mexico)

    1998-07-01

    The Boiling Water nuclear Reactors (BWR) have presented stress corrosion problems, mainly in components and pipes of the primary system, provoking negative impacts in the performance of energy generator plants, as well as the increasing in the radiation exposure to personnel involucred. This problem has caused development of research programs, which are guided to find solution alternatives for the phenomena control. Among results of greater relevance the control for the reactor water chemistry stands out particularly in the impurities concentration and oxidation of radiolysis products; as well as the supervision in the materials selection and the stresses levels reduction. The present work presents the methods which can be applied to diminish the problems of stress corrosion in BWR reactors. (Author)

  19. 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)

  20. Reactor pressure vessel thermal annealing

    International Nuclear Information System (INIS)

    Lee, A.D.

    1997-01-01

    The steel plates and/or forgings and welds in the beltline region of a reactor pressure vessel (RPV) are subject to embrittlement from neutron irradiation. This embrittlement causes the fracture toughness of the beltline materials to be less than the fracture toughness of the unirradiated material. Material properties of RPVs that have been irradiated and embrittled are recoverable through thermal annealing of the vessel. The amount of recovery primarily depends on the level of the irradiation embrittlement, the chemical composition of the steel, and the annealing temperature and time. Since annealing is an option for extending the service lives of RPVs or establishing less restrictive pressure-temperature (P-T) limits; the industry, the Department of Energy (DOE) and the Nuclear Regulatory Commission (NRC) have assisted in efforts to determine the viability of thermal annealing for embrittlement recovery. General guidance for in-service annealing is provided in American Society for Testing and Materials (ASTM) Standard E 509-86. In addition, the American Society of Mechanical Engineers (ASME) Code Case N-557 addresses annealing conditions (temperature and duration), temperature monitoring, evaluation of loadings, and non-destructive examination techniques. The NRC thermal annealing rule (10 CFR 50.66) was approved by the Commission and published in the Federal Register on December 19, 1995. The Regulatory Guide on thermal annealing (RG 1.162) was processed in parallel with the rule package and was published on February 15, 1996. RG 1.162 contains a listing of issues that need to be addressed for thermal annealing of an RPV. The RG also provides alternatives for predicting re-embrittlement trends after the thermal anneal has been completed. This paper gives an overview of methodology and recent technical references that are associated with thermal annealing. Results from the DOE annealing prototype demonstration project, as well as NRC activities related to the

  1. Simulations of ex-vessel fuel coolant interactions in a Nordic BWR using MC3D code

    International Nuclear Information System (INIS)

    Thakre, S.; Ma, W.

    2013-08-01

    Nordic Boiling Water Reactors (BWRs) employ a drywell cavity flooding technique as a nuclear severe accident management strategy. In case of core melt accident where the reactor pressure vessel will fail and the melt will eject from the lower head and fall into a water pool, may be in the form of a continuous jet. It is assumed that the melt jet will fragment, quench and form a coolable debris bed into the water pool. The melt interaction with a water pool may cause an energetic steam explosion which creates a potential risk towards the integrity of containment, leading to fission products release into the atmosphere. The results of the APRI-7 project suggest that the significant damage to containment structures by steam explosion cannot be ruled according to the state-of-the-art knowledge about corresponding accident scenario. In the follow-up project APRI-8 (2012-2016) one of the goals of the KTH research is to resolve the steam explosion energetics (SEE) issue, developing a risk-oriented framework for quantifying conditional threats to containment integrity for a Nordic type BWR. The present study deals with the premixing and explosion phase calculations of a Nordic BWR dry cavity, using MC3D, a multiphase CFD code for fuel coolant interactions. The main goal of the study is the assessment of pressure buildup in the cavity and the impact loading on the side walls. The conditions for the calculations are used from the SERENA-II BWR case exercise. The other objective was to do the sensitivity analysis of the parameters in modeling of fuel coolant interactions, which can help to reduce uncertainty in assessment of steam explosion energetics. The results show that the amount of liquid melt droplets in the water (region of void<0.6) is maximum even before reaching the jet at the bottom. In the explosion phase, maximum pressure is attained at the bottom and the maximum impulse on the wall is at the bottom of the wall. The analysis is carried out using two different

  2. Simulations of ex-vessel fuel coolant interactions in a Nordic BWR using MC3D code

    Energy Technology Data Exchange (ETDEWEB)

    Thakre, S.; Ma, W. [Royal Institute of Technology, KTH. Div. of Nuclear Power Safety, Stockholm (Sweden)

    2013-08-15

    Nordic Boiling Water Reactors (BWRs) employ a drywell cavity flooding technique as a nuclear severe accident management strategy. In case of core melt accident where the reactor pressure vessel will fail and the melt will eject from the lower head and fall into a water pool, may be in the form of a continuous jet. It is assumed that the melt jet will fragment, quench and form a coolable debris bed into the water pool. The melt interaction with a water pool may cause an energetic steam explosion which creates a potential risk towards the integrity of containment, leading to fission products release into the atmosphere. The results of the APRI-7 project suggest that the significant damage to containment structures by steam explosion cannot be ruled according to the state-of-the-art knowledge about corresponding accident scenario. In the follow-up project APRI-8 (2012-2016) one of the goals of the KTH research is to resolve the steam explosion energetics (SEE) issue, developing a risk-oriented framework for quantifying conditional threats to containment integrity for a Nordic type BWR. The present study deals with the premixing and explosion phase calculations of a Nordic BWR dry cavity, using MC3D, a multiphase CFD code for fuel coolant interactions. The main goal of the study is the assessment of pressure buildup in the cavity and the impact loading on the side walls. The conditions for the calculations are used from the SERENA-II BWR case exercise. The other objective was to do the sensitivity analysis of the parameters in modeling of fuel coolant interactions, which can help to reduce uncertainty in assessment of steam explosion energetics. The results show that the amount of liquid melt droplets in the water (region of void<0.6) is maximum even before reaching the jet at the bottom. In the explosion phase, maximum pressure is attained at the bottom and the maximum impulse on the wall is at the bottom of the wall. The analysis is carried out using two different

  3. Repair and preventive maintenance technology for BWR reactor internals and piping

    International Nuclear Information System (INIS)

    Ootsubo, Tooru; Itou, Takashi; Sakashita, Akihiro

    2009-01-01

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

  4. Reactor pressure vessel. Status report

    International Nuclear Information System (INIS)

    Elliot, B.J.; Hackett, E.M.; Lee, A.D.

    1996-10-01

    This report describes the issues raised as a result of the staffs review of Generic Letter (GL) 92-01, Revision 1, responses and plant-specific reactor pressure vessel (RPV) assessments and the actions taken or work in progress to address these issues. In addition, the report describes actions taken by the staff and the nuclear industry to develop a thermal annealing process for use at U.S. commercial nuclear power plants. This process is intended to be used as a means of mitigating the effects of neutron radiation on the fracture toughness of RPV materials. The Nuclear Regulatory Commission (NRC) issued GL 92-01, Revision 1, Supplement 1, to obtain information needed to assess compliance with regulatory requirements and licensee commitments regarding RPV integrity. GL 92-01, Revision 1, Supplement 1, was issued as a result of generic issues that were raised in the NRC staff's reviews of licensee responses to GL 92-01, Revision 1, and plant-specific RPV evaluations. In particular, an integrated review of all data submitted in response to GL 92-01, Revision 1, indicated that licensees may not have considered all relevant data in their RPV assessments. This report is representative of submittals to and evaluations by the staff as of September 30, 1996. An update of this report will be issued at a later date

  5. 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)

  6. 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

  7. 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.

  8. Recent trends in the mitigation of the IGSCC through modifications in the water chemistry of BWR reactors; Tendencias recientes en la mitigacion del IGSCC mediante modificaciones en la quimica del agua de reactores BWR

    Energy Technology Data Exchange (ETDEWEB)

    Diaz S, A.; Robles, E.F. [ININ, 52045 Ocoyoacac, Estado de Mexico (Mexico)

    2003-07-01

    During the last years, the Nuclear Power stations had been that to adequate or to modify the parameters and operational conditions, attempting to maintain and to safeguard the integrity and functionality of its components and systems, as well as the personnel safety involved in its operation. In a Boiling water reactor (BWR), the chemical control of the water, constitutes one of the fundamental aspects to get a sure and reliable operation, having as main objectives: (a) The protection of the reactor vessel, of the structural materials of the same one and of the pipes and components of those recirculation systems against the Intergranular stress corrosion phenomena (IGSCC); (b) To guarantee the integrity of the nuclear fuel minimizing the corrosion phenomena in the fuel elements; and (c) The reduction of the operational dose of the personnel involved directly in the operation and maintenance by means of the control of the activated corrosion products. (Author)

  9. Simulation of the automatic depressurization system (Ads) for a boiling water reactor (BWR) based on RELAP; Simulacion del sistema de despresurizacion automatica (ADS) para un reactor de agua en ebullicion (BWR) basado en RELAP

    Energy Technology Data Exchange (ETDEWEB)

    Ramirez G, C.; Chavez M, C., E-mail: ces.raga@gmail.com [UNAM, Facultad de Ingenieria, Circuito Interior, Ciudad Universitaria, 04510 Mexico D. F. (Mexico)

    2012-10-15

    The automatic depressurization system (Ads) of the boiling water reactor (BWR) like part of the emergency cooling systems is designed to liberate the vapor pressure of the reactor vessel, as well as the main vapor lines. At the present time in the Engineering Faculty, UNAM personnel works in the simulation of the Laguna Verde reactor based on the nuclear code RELAP/SCADAP and in the incorporation to the same of the emergency cooling systems. The simulation of the emergency cooling systems began with the inclusion of two hydrodynamic volumes, one source and another drain, and the incorporation of the initiation logic for each emergency system. In this work is defined and designed a simplified model of Ads of the reactor, considering a detail level based on the main elements that compose it. As tool to implement the proposed model, the RELAP code was used. The simulated main functions of Ads are centered in the quick depressurization of the reactor by means of the vapor discharge through the relief/safety valves to the suppression pool, and, in the event of break of the main vapor line, the reduction of the vessel pressure operates for that the cooling systems of the core to low pressure (Lpcs and Lpci) they can begin their operation. (Author)

  10. Modelling perspectives on radiation chemistry in BWR reactor core

    International Nuclear Information System (INIS)

    Ibe, Eishi

    1991-01-01

    Development of a full-system boiling water reactor core model started in 1982. The model included a two-region reactor core, one with and one without boiling. Key design parameters consider variable dose rates in a three-layer liquid downcomer. Dose rates in the core and downcomer include both generation and recombination reactions of species. Agreement is good between calculations and experimental data of oxygen concentration as a function of hydrogen concentration for different bubble sizes. Oxygen concentration is reduced in the reactor pressure vessel (RPV) by increasing bubble size. The multilayer model follows the oxygen data better than a single-layered model at high concentrations of hydrogen. Key reactions are reduced to five radiolysis reactions and four decomposition reactions for hydrogen peroxide. Calculations by the DOT 3 code showed dose rates from neutrons and gamma rays in various parts of the core. Concentrations of oxygen, hydrogen peroxide, and hydrogen were calculated by the model as a function of time from core inlet. Similar calculations for NWC and HWC were made as a function of height from core inlet both in the boiling channel an the bypass channel. Finally the model was applied to calculate the oxygen plus half the hydrogen peroxide concentrations as a function of hydrogen concentration to compare with data from five plants. Power density distribution with core height was given for an early stage and an end stage of a cycle. Increases of dose rates in the turbine for seven plants were shown as a function of increased hydrogen concentration in the reactor water

  11. Propagation of cracks by stress corrosion in conditions of BWR type reactor; Propagacion de grietas por corrosion bajo esfuerzo en condiciones de reactor de agua en ebullicion (BWR)

    Energy Technology Data Exchange (ETDEWEB)

    Merino C, F.J. [ININ, 52045 Estado de Mexico (Mexico); Fuentes C, P. [ITT, Metepec, Estado de Mexico (Mexico)]. E-mail: fjmc@nuclear.inin.mx

    2004-07-01

    In this work, the obtained results when applying the Hydrogen Chemistry to a test tube type Compact Tension (CT), built in austenitic stainless steel 304l, simulating the conditions to those that it operates a Boiling Water Reactor (BWR), temperature 288 C and pressure of 8 MPa are presented. With the application of this water chemistry, seeks to be proven the diminution of the crack propagation speed. (Author)

  12. Study of the Utilization BWR Type Nuclear Power Reactor for Desalination Process

    International Nuclear Information System (INIS)

    Itjeu Karliana; Sumijanto; Dhandhang Purwadi, M.

    2008-01-01

    The needs of fresh water increased by rapid population growth and industrials expansion, but these demands can not be prepared naturally. Following this case, seawater desalination becomes the primer option which can fulfill the need through the nuclear desalination technology. The coupled nuclear power reactor enables to supply thermal energy for auxiliary equipment and pumps operation. The utilization study of power reactor type BWR coupled with desalination process has been performed. The goal of study is to obtain characteristic data of desalted water specification which desalination system coupling with nuclear power plant produced energy for desalination process. The study is carried out by browsing data and information, and comprehensive review of thermal energy correlation between NPP with desalination process installation. According to reviewing are found that the thermal energy and electric power utilization from the nuclear power reactor are enable to remove the seawater to produce desalted water and also to operate auxiliary equipments. The assessment results is VK-300 reactor prototype, BWR type 250 MW(e) power are cogeneration unit can supplied hot steam temperature 285 °C to the extraction turbine to empower 150 MW electric power, and a part of hot steam 130 °C is use to operate desalination process and remind heat is distribute to the municipal and offices at that region. The coupled of VK-300 reactor power type BWR with desalination installation of MED type enable to produce desalted water with high quality distillate. Based on the economic calculation that the VK-300 reactor power of BWR type produced water distillate capacity is 300.000 m 3 /hour with cost US$ 0.58/m 3 . The coupling VK-300 reactor power type BWR with MED desalination plant is competitive economically. (author)

  13. Recent trends in the mitigation of the IGSCC through modifications in the water chemistry of BWR reactors

    International Nuclear Information System (INIS)

    Diaz S, A.; Robles, E.F.

    2003-01-01

    During the last years, the Nuclear Power stations had been that to adequate or to modify the parameters and operational conditions, attempting to maintain and to safeguard the integrity and functionality of its components and systems, as well as the personnel safety involved in its operation. In a Boiling water reactor (BWR), the chemical control of the water, constitutes one of the fundamental aspects to get a sure and reliable operation, having as main objectives: (a) The protection of the reactor vessel, of the structural materials of the same one and of the pipes and components of those recirculation systems against the Intergranular stress corrosion phenomena (IGSCC); (b) To guarantee the integrity of the nuclear fuel minimizing the corrosion phenomena in the fuel elements; and (c) The reduction of the operational dose of the personnel involved directly in the operation and maintenance by means of the control of the activated corrosion products. (Author)

  14. Cracks propagation by stress corrosion cracking in conditions of Boiling Water Reactor (BWR); Propagacion de grietas por corrosion bajo esfuerzo en condiciones de reactor de agua hirviente (BWR)

    Energy Technology Data Exchange (ETDEWEB)

    Fuentes C, P

    2003-07-01

    This work presents the results of the assays carried out in the Laboratory of Hot Cells of the National Institute of Nuclear Research (ININ) to a type test tube Compact Tension (CT), built in steel austenitic stainless type 304L, simulating those conditions those that it operates a Boiling Water Reactor (BWR), at temperature 288 C and pressure of 8 MPa, to determine the speed to which the cracks spread in this material that is of the one that different components of a reactor are made, among those that it highlights the reactor core vessel. The application of the Hydrogen Chemistry of the Water is presented (HWC) that is one alternative to diminish the corrosion effect low stress in the component, this is gets controlling the quantity of oxygen and of hydrogen as well as the conductivity of the water. The rehearsal is made following the principles of the Mechanics of Elastic Lineal Fracture (LEFM) that considers a crack of defined size with little plastic deformation in the tip of this; the measurement of crack advance is continued with the technique of potential drop of direct current of alternating signal, this is contained inside the standard Astm E-647 (Method of Test Standard for the Measurement of Speed of Growth of Crack by fatigue) that is the one that indicates us as carrying out this test. The specifications that should complete the test tubes that are rehearsed as for their dimensions, it forms, finish and determination of mechanical properties (tenacity to the fracture mainly) they are contained inside the norm Astm E-399, the one which it is also based on the principles of the fracture mechanics. The obtained results were part of a database to be compared with those of other rehearsals under different conditions, Normal Chemistry of the Water (NWC) and it dilutes with high content of O{sub 2}; to determine the conditions that slow more the phenomena of stress corrosion cracking, as well as the effectiveness of the used chemistry and of the method of

  15. 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

  16. 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

  17. 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)

  18. Boron concentration evolution in the temporary curtains of a BWR reactor. Burcur code

    International Nuclear Information System (INIS)

    Cano Aguado, M.; Perlado Martin, J.M.; Minguez Torres, E.

    1977-01-01

    The theoretical model and the user's guide of the code Burcur is included. This code analyzes the burnable poison concentration of the temporary curtains as a function of time, for BWR reactors of the 7 x 7 design. The computing time being reasonably short, the number of burnup steps is as high as necessary.(author) [es

  19. 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.)

  20. 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

  1. 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

  2. Accident sequence analysis for a BWR [Boiling Water Reactor] during low power and shutdown operations

    International Nuclear Information System (INIS)

    Whitehead, D.W.; Hake, T.M.

    1990-01-01

    Most previous Probabilistic Risk Assessments have excluded consideration of accidents initiated in low power and shutdown modes of operation. A study of the risk associated with operation in low power and shutdown is being performed at Sandia National Laboratories for a US Boiling Water Reactor (BWR). This paper describes the proposed methodology for the analysis of the risk associated with the operation of a BWR during low power and shutdown modes and presents preliminary information resulting from the application of the methodology. 2 refs., 2 tabs

  3. 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

  4. Fast-neutron nuclear reactor vessel

    International Nuclear Information System (INIS)

    Presciuttini, L.

    1984-01-01

    The reactor vessel comprises a cylindrical shell, of which axis is vertical, coupled at its lower part with a dished bottom. The reactor core rests on a support plate bearing on the bottom of the vessel. The above dished bottom comprises a spherical sector having the same radius and the same axis as the cylindrical shell and joining the lower part of the shell, and a spherical head of which radius is a little more important than the spherical sector one. A cylindrical support for the reactor core is attached to the vessel at the joint between the two dished sections. The invention applies more particularly to integrated type reactors cooled by liquid sodium [fr

  5. Preliminary Assessment of the Possible BWR Core/Vessel Damage States for Fukushima Daiichi Station Blackout Scenarios Using RELAP/SCDAPSIM

    Directory of Open Access Journals (Sweden)

    C. M. Allison

    2012-01-01

    Full Text Available Immediately after the accident at Fukushima Daiichi, Innovative Systems Software and other members of the international SCDAP Development and Training Program started an assessment of the possible core/vessel damage states of the Fukushima Daiichi Units 1–3. The assessment included a brief review of relevant severe accident experiments and a series of detailed calculations using RELAP/SCDAPSIM. The calculations used a detailed RELAP/SCDAPSIM model of the Laguna Verde BWR vessel and related reactor cooling systems. The Laguna Verde models were provided by the Comision Nacional de Seguridad Nuclear y Salvaguardias, the Mexican nuclear regulatory authority. The initial assessment was originally presented to the International Atomic Energy Agency on March 21 to support their emergency response team and later to our Japanese members to support their Fukushima Daiichi specific analysis and model development.

  6. Crack growth behaviour of low-alloy steels for pressure boundary components under transient light water reactor operating conditions - CASTOC, Part I: BWR/NWC conditions

    International Nuclear Information System (INIS)

    Ritter, S.; Seifert, H.P.; Devrient, B.; Roth, A.; Ehrnsten, U.; Ernestova, M.; Zamboch, M.; Foehl, J.; Weissenberg, T.; Gomez-Briceno, D.; Lapena, J.

    2004-01-01

    One of the ageing phenomena of pressure boundary components of light water reactors (LWR) is environmentally-assisted cracking (EAC). The project CASTOC (5. Framework Programme of the EU) was launched September 2000 with six European partners and terminated August 2003. It was focused in particular on the EAC behaviour of low-alloy steels (LAS) and to some extent to weld metal, heat affected zone and the influence of an austenitic cladding. The main objective was directed to the clarification of EAC crack growth behaviour/mechanism of LAS in high-temperature water under steady-state power operation (constant load) and transient operating conditions (e.g., start-up/shut-down, transients in water chemistry and load). Autoclave tests were performed with Western and Russian type reactor pressure vessel steels under simulated boiling water reactor (BWR)/normal water chemistry (NWC) and pressurised water reactor (VVER) conditions. The investigations were performed with fracture mechanics specimens of different sizes and geometries. The applied loading comprised cyclic loads, static loads and load spectra where the static load was periodically interrupted by partial unloading. With regard to water chemistry, the oxygen content (VVER) and impurities of sulphate and chlorides (BWR) were varied beyond allowable limits for continuous operation. The current paper summarises the most important crack growth results obtained under simulated BWR/NWC conditions. The results are discussed in the context of the current crack growth rate curves in the corresponding nuclear codes. (authors)

  7. Crack growth behaviour of low-alloy steels for pressure boundary components under transient light water reactor operating conditions - CASTOC, Part I: BWR/NWC conditions

    Energy Technology Data Exchange (ETDEWEB)

    Ritter, S.; Seifert, H.P. [Paul Scherrer Institute, PSI, Villigen (Switzerland); Devrient, B.; Roth, A. [Framatome ANP GmbH, Erlangen (Germany); Ehrnsten, U. [VTT Industrial Systems, Espoo (Finland); Ernestova, M.; Zamboch, M. [Nuclear Research Institute, NRI, Rez (Czech Republic); Foehl, J.; Weissenberg, T. [Staatliche Materialpruefungsanstalt, MPA, Stuttgart (Germany); Gomez-Briceno, D.; Lapena, J. [Centro de Investigaciones Energeticas Medioambientales y Tecnologicas, CIEMAT, Madrid (Spain)

    2004-07-01

    One of the ageing phenomena of pressure boundary components of light water reactors (LWR) is environmentally-assisted cracking (EAC). The project CASTOC (5. Framework Programme of the EU) was launched September 2000 with six European partners and terminated August 2003. It was focused in particular on the EAC behaviour of low-alloy steels (LAS) and to some extent to weld metal, heat affected zone and the influence of an austenitic cladding. The main objective was directed to the clarification of EAC crack growth behaviour/mechanism of LAS in high-temperature water under steady-state power operation (constant load) and transient operating conditions (e.g., start-up/shut-down, transients in water chemistry and load). Autoclave tests were performed with Western and Russian type reactor pressure vessel steels under simulated boiling water reactor (BWR)/normal water chemistry (NWC) and pressurised water reactor (VVER) conditions. The investigations were performed with fracture mechanics specimens of different sizes and geometries. The applied loading comprised cyclic loads, static loads and load spectra where the static load was periodically interrupted by partial unloading. With regard to water chemistry, the oxygen content (VVER) and impurities of sulphate and chlorides (BWR) were varied beyond allowable limits for continuous operation. The current paper summarises the most important crack growth results obtained under simulated BWR/NWC conditions. The results are discussed in the context of the current crack growth rate curves in the corresponding nuclear codes. (authors)

  8. Applied methods for mitigation of damage by stress corrosion in BWR type reactors

    International Nuclear Information System (INIS)

    Hernandez C, R.; Diaz S, A.; Gachuz M, M.; Arganis J, C.

    1998-01-01

    The Boiling Water nuclear Reactors (BWR) have presented stress corrosion problems, mainly in components and pipes of the primary system, provoking negative impacts in the performance of energy generator plants, as well as the increasing in the radiation exposure to personnel involucred. This problem has caused development of research programs, which are guided to find solution alternatives for the phenomena control. Among results of greater relevance the control for the reactor water chemistry stands out particularly in the impurities concentration and oxidation of radiolysis products; as well as the supervision in the materials selection and the stresses levels reduction. The present work presents the methods which can be applied to diminish the problems of stress corrosion in BWR reactors. (Author)

  9. Fuel loading method to exchangeable reactor core of BWR type reactor and its core

    International Nuclear Information System (INIS)

    Koguchi, Kazushige.

    1995-01-01

    In a fuel loading method for an exchangeable reactor core of a BWR type reactor, at least two kinds of fresh fuel assemblies having different reactivities between axial upper and lower portions are preliminarily prepared, and upon taking out fuel assemblies of advanced combustion and loading the fresh fuel assemblies dispersingly, they are disposed so as to attain a predetermined axial power distribution in the reactor. At least two kinds of fresh fuel assemblies have a content of burnable poisons different between the axial upper portion and lower portions. In addition, reactivity characteristics are made different at a region higher than the central boundary and a region lower than the central boundary which is set within a range of about 6/24 to 16/24 from the lower portion of the fuel effective length. There can be attained axial power distribution as desired such as easy optimization of the axial power distribution, high flexibility, and flexible flattening of the power distribution, and it requires no special change in view of the design and has a good economical property. (N.H.)

  10. Pressurized water reactor with reactor pressure vessel

    International Nuclear Information System (INIS)

    Werres, L.

    1985-01-01

    The pressure vessel has a cylindrical jacket with a domed floor. A guide is arranged on the domed floor to even out the flow in the core. It consists of a cylindrical jacket, whose lower end has slots and fins. These fins are welded to the domed floor. (orig./PW)

  11. Pressurized water reactor with reactor pressure vessel

    International Nuclear Information System (INIS)

    Werres, L.

    1980-01-01

    The pressure vessel has a cylindrical jacket with a domed floor. A guide is arranged on the domed floor to even out the flow in the core. It consists of a cylindrical jacket, whose lower end has slots and fins. These fins are welded to the domed floor. (DG) [de

  12. Generic BWR-4 degraded core in-vessel study. Status report

    International Nuclear Information System (INIS)

    1984-11-01

    Original intent of this project was to produce a phenomenological study of the in-vessel degradation which occurs during the TQUX and TQUV sequences for a generic BWR-4 from the initiation of the FSAR Chapter 15 operational transient through core debris bed formation to the failure of the primary pressure boundary. Bounding calculations were to be performed for the two high pressure and low pressure non-LOCA scenarios to assess the uncertainties in the current state of knowledge regarding the source terms for containment integrity studies. Source terms as such were defined in terms of hydrogen generation, unreacted metal, and coolant inventroy, and in terms of the form, sequencing and mode of dispersal through the primary vessel boundary. Fission product release was not to be considered as part of this study. Premature termination of the project, however, led to the dicontinuation of work on an as is basis. Work on the in-core phase from the point of scram to core debris bed formation was largely completed. A preliminary scoping calculation on the debris bed phase had been initiated. This report documents the status of the study at termination

  13. Development of new irradiation facility for BWR safety research

    International Nuclear Information System (INIS)

    Okada, Yuji; Magome, Hirokatsu; Iida, Kazuhiro; Hanawa, Hiroshi; Ohmi, Masao

    2013-01-01

    In JAEA (Japan Atomic Energy Agency), about the irradiation embrittlement of the reactor pressure vessel and the stress corrosion cracking of reactor core composition apparatus concerning the long-term use of the light water reactor (BWR), in order to check the influence of the temperature, pressure, and water quality, etc on BWR condition. The water environmental control facility which performs irradiation assisted stress corrosion-cracking (IASCC) evaluation under BWR irradiation environment was fabricated in JMTR (Japan Materials Testing Reactor). This report is described the outline of manufacture of the water environmental control facility for doing an irradiation test using the saturation temperature capsule after JMTR re-operation. (author)

  14. Connected analysis nuclear-thermo-hydraulic of parallel channels of a BWR reactor using distributed computation; Analisis acoplado nuclear-termohidraulico de canales paralelos de un reactor BWR empleando computacion distribuida

    Energy Technology Data Exchange (ETDEWEB)

    Campos Gonzalez, Rina Margarita

    2007-07-15

    This work consists of the integration of three models previously developed which are described widely in Literature: model of the thermo-hydraulic channel, model of the modal neutronic and the model of the recirculation bows. The tool used for this connection of models is the PVM system, Parallel Virtual Machine that allowed paralleling the model by means of the concept of distributed computation. The purpose of making this connection of models is the one of obtaining a more complete tool than better represents the real configuration and the phenomenology of the nucleus of a BWR reactor, thus obtaining better results. In addition to maintaining the flexibility to improve the resulting model at any time, since the very complex or sophisticated models are difficult to improve being impossible to modify the equations they use and can include variables that are not of primary importance in the tackled problem or that mask relations among variables due to the excess of results. Also maintaining the flexibility for adding component of models or systems of the BWR reactor, all of this following the modeling needs. The Swedish Ringhals power plant was chosen to characterize the resulting connected model for counting on a Stability Benchmark that offers the opportunity to count on real plant data. Besides that in case 9 of cycle 14 of this Benchamark oscillations outside phase appeared, which are from great interest because the detection systems that register the average of the power of the nucleus do not detect them. Additionally in this work the model of the recirculation bows as an independent module is obtained in an individual way, since this model belongs to another work and works connected to the reactor vessel. The model of the recirculation bows is able to model several transients of interest, as it is shown in the Appendix A of this work, among which are found the tripping of recirculation pumps or the transference at low or high velocity of them. The scope of the

  15. 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

  16. 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.)

  17. 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.)

  18. 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)

  19. Core followup studies of the Tarapur Reactors with the three dimensional BWR simulator COMTEG

    Energy Technology Data Exchange (ETDEWEB)

    Dwivedi, S. R.; Jagannathan, V.; Mohanakrishnan, P.; Srinivasan, K. R.; Rastogi, B. P.

    1976-07-01

    Both the units of the Tarapur Atomic Power Station started operation in the year 1969. Since then, these units have completed three cycles. For efficient operation and fuel management of these reactors, a three dimensional BWR simulator COMETG has been developed. The reactors are closely being followed using the simulator. The detailed analyses for cycle 3/4 operation of both the units are described in the paper. The results show very good agreement between calculated and measured values. It is concluded that reactor core behaviour could be predicted in a satisfactory manner with the core simulator COMETG.

  20. The design and use of proficiency based BWR reactor maintenance and refuelling training mockups

    International Nuclear Information System (INIS)

    Ford, G.J.

    1996-01-01

    The purpose of this paper is to describe the ABB experience with the design and use of boiling water reactor training facilities. The training programs were developed and implemented in cooperation with the nuclear utilities. ABB operates two facilities, the ABB ATOM Light Water Reactor Service Center located in Vasteras, Sweden, and the ABB Combustion Engineering Nuclear Operations BWR Training Center located in Chattanooga, Tennessee, USA. The focus of the training centers are reactor maintenance and refueling activities plus the capability to develop and qualify tools, procedures and repair techniques

  1. 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.

  2. 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

  3. Application of the Synthesis method to the calculations of neutron flow in 3D in the enveloping of a BWR reactor with the DORT code

    International Nuclear Information System (INIS)

    Xolocostli M, J.V.; Gomez T, A.M.; Palacios H, J.C.

    2006-01-01

    The surveillance program of the vessel materials of a BWR reactor requires the determination of the neutron flux in 3D in the core enveloping. To carry out these calculations of the neutron flux, the Regulatory Guide 1.190 of the NRC recommends the use of the following codes: MCNP, TORT and DORT. In the case of using the DORT code, the one which solves the transport equation in discreet coordinates and in two dimensions (xy, rθ, and rz), the regulatory guide in reference, requires to make an approach of the flow in three dimensions by means of the call Synthesis Method. It is denominated like this due to that a flow representation in 3D is achieved 'combining' or 'synthesizing' the calculated flows by DORT in rθ, rz and r. In this work the application of the Synthesis Method it is presented, according to the Regulatory Guide 1.190, to determine the 3D flows in a BWR reactor. To achieve the above mentioned it was implemented the Synthesis Method in a computer program developed in the ININ to which is denominated SYNTHESIS. This program applies the synthesis method, and is 'coupled' with the DORT code to determine by this way the neutronic fluxes in 3D on the enveloping of a BWR reactor. (Author)

  4. 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)

  5. Control in fabrication of PWR and BWR type reactor fuel elements

    International Nuclear Information System (INIS)

    Gorskij, V.V.

    1981-01-01

    Both destructive and non-destructive testing methods now in use in fabrication of BWR and PWR type reactor fuel elements at foreign plants are reviewed. Technological procedures applied in fabrication of fuel elements and fuel assemblies are described. Major attention is paid to radiographic, ultrasonic, metallographic, visual and autoclavic testings. A correspondence of the methods applied to the ASTM standards is discussed. The most part of the countries are concluded the apply similar testing methods enabling one to reliably evaluate the quality of primary materials and fabricated fuel elements and thus meeting the demands to contemporary PWR and BWR type reactor fuel elements. Practically all fuel element and pipe fabrication plants in Western Europe, Asia and America use the ASTM standards as the basis for the quality contr [ru

  6. Kinematics of two-phase mixture level motion in BWR pressure vessels

    International Nuclear Information System (INIS)

    Wulff, W.; Cheng, H.S.; Mallen, A.N.; Stritar, A.

    1985-01-01

    A model is presented for predicting two-phase mixture level elevations in BWR systems. The model accounts for the particular geometry and conditions in a BWR system during Small-Break Loss of Coolant Accidents. The model presented here is particularly suitable for efficient, high-speed simulations on small minicomputers. The model has been implemented and tested. Results are shown from BWR ATWS simulations

  7. Propagation of cracks by stress corrosion in conditions of BWR type reactor

    International Nuclear Information System (INIS)

    Merino C, F.J.; Fuentes C, P.

    2004-01-01

    In this work, the obtained results when applying the Hydrogen Chemistry to a test tube type Compact Tension (CT), built in austenitic stainless steel 304l, simulating the conditions to those that it operates a Boiling Water Reactor (BWR), temperature 288 C and pressure of 8 MPa are presented. With the application of this water chemistry, seeks to be proven the diminution of the crack propagation speed. (Author)

  8. Fracture assessment of a main reactor coolant pump in a BWR with encountered defects

    Energy Technology Data Exchange (ETDEWEB)

    Brickstad, B [Swedish Plant Inspectorate, Stockholm (Sweden)

    1988-12-31

    This document presents a case-study fracture assessment in BWR type reactor components. A cast stainless steel presenting defects due to thermal is studied. The stress analysis performed by aid of a finite element technique shows that a Leak Before Break situation could be expected. Nevertheless, it may be concluded that the cross section of the pump where the defect area was located can withstand very deep cracks before the risk of failure becomes unacceptable. (TEC).

  9. 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)

  10. Inspection accessibility study of the Millstones Unit 1 Reactor Pressure Vessel

    International Nuclear Information System (INIS)

    Calhoun, G.; Kapoor, A.; Davis, J.B.

    1990-07-01

    The need for more extensive ultrasonic examination of the Reactor Pressure Vessel (RPV) shell welds at some of the BWR plants is becoming apparent. Recent NRC comments regarding limits on future inspection relief requests and proposed revisions to the ASME Boiler and Pressure Vessel Code, Section 11 have caused BWR plant owners and vendors of inspection services to look at inspection tooling that would allow for the ultrasonic examination of RPV shell welds previously considered inaccessible. One approach to inspection of the RPV shell welds that will allow for greater coverage at some BWR plants is to perform these inspections from inside the vessel. In late 1988, Westinghouse received a contract from Northeast Utilities to examine the flange ligament areas and the shell welds in the upper portion of the Millstone Unit 1 PRV during an outage scheduled to begin in April 1989. This examination was performed with an inspection tool placed inside of the vessel and the UDRPS data acquisition system. The inspection tool was based on tooling used routinely at the Swedish and Finnish BWRs. The object of RP C105-1 was to use the information gathered prior to and during the Millstone Unit 1 examination and to perform an inspection accessibility study and prepare a report covering the lessons learned during the examination that would benefit other BWR utilities considering similar inspections. The final report consists of two volumes. This document, Volume 1, describes a 3D model of the Millstone Unit 1 PRV that was based on information obtained from plant drawings. The model is a useful tool for visualizing areas of limited access from both inside and outside of the RPV. 5 refs., 38 figs., 4 tabs

  11. 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

  12. BWR type nuclear power plant

    International Nuclear Information System (INIS)

    Matsumoto, Kosuke.

    1991-01-01

    In a BWR type nuclear power plant in which reactor water in a reactor pressure vessel can be drained to a waste processing system by way of reactor recycling pipeways and remaining heat removal system pipeways, a pressurized air supply device is disposed for supplying air for pressurizing reactor water to the inside of the reactor pressure vessel by way of an upper head. With such a constitution, since the pressurized air sent from the pressurized air supply device above the reactor pressure vessel for the reactor water discharging pressure upon draining, the water draining pressure is increased compared with a conventional case and, accordingly, the amount of drained water is not reduced even in the latter half of draining. Accordingly, the draining efficiency can be improved and only a relatively short period of time is required till the completion of the draining, which can improve safety and save labors. (T.M.)

  13. Radionuclide inventory calculation in VVER and BWR reactor

    International Nuclear Information System (INIS)

    Bouhaddane, A.; Farkas, F.; Slugen, V.; Ackermann, L.; Schienbein, M.

    2014-01-01

    The paper shows different aspects in the radionuclide inventory determination. Precise determination of the neutron flux distribution, presented for a BRW reactor, is vital for the activation calculations. The precision can be improved utilizing variance reduction methods as importance treatment, weight windows etc. Direct calculation of the radionuclide inventory via Monte Carlo code is presented for a VVER reactor. Burn-up option utilized in this calculation appears to be proper for reactor internal components. However, it will not be probably effective outside the reactor core. Further calculations in this area are required to support the forth-set findings. (authors)

  14. Cracks propagation by stress corrosion cracking in conditions of Boiling Water Reactor (BWR)

    International Nuclear Information System (INIS)

    Fuentes C, P.

    2003-01-01

    This work presents the results of the assays carried out in the Laboratory of Hot Cells of the National Institute of Nuclear Research (ININ) to a type test tube Compact Tension (CT), built in steel austenitic stainless type 304L, simulating those conditions those that it operates a Boiling Water Reactor (BWR), at temperature 288 C and pressure of 8 MPa, to determine the speed to which the cracks spread in this material that is of the one that different components of a reactor are made, among those that it highlights the reactor core vessel. The application of the Hydrogen Chemistry of the Water is presented (HWC) that is one alternative to diminish the corrosion effect low stress in the component, this is gets controlling the quantity of oxygen and of hydrogen as well as the conductivity of the water. The rehearsal is made following the principles of the Mechanics of Elastic Lineal Fracture (LEFM) that considers a crack of defined size with little plastic deformation in the tip of this; the measurement of crack advance is continued with the technique of potential drop of direct current of alternating signal, this is contained inside the standard Astm E-647 (Method of Test Standard for the Measurement of Speed of Growth of Crack by fatigue) that is the one that indicates us as carrying out this test. The specifications that should complete the test tubes that are rehearsed as for their dimensions, it forms, finish and determination of mechanical properties (tenacity to the fracture mainly) they are contained inside the norm Astm E-399, the one which it is also based on the principles of the fracture mechanics. The obtained results were part of a database to be compared with those of other rehearsals under different conditions, Normal Chemistry of the Water (NWC) and it dilutes with high content of O 2 ; to determine the conditions that slow more the phenomena of stress corrosion cracking, as well as the effectiveness of the used chemistry and of the method of

  15. 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

  16. Simulation of hydrogen deflagration and detonation in a BWR reactor building

    International Nuclear Information System (INIS)

    Manninen, M.; Silde, A.; Lindholm, I.; Huhtanen, R.; Sjoevall, H.

    2002-01-01

    A systematic study was carried out to investigate the hydrogen behaviour in a BWR reactor building during a severe accident. BWR core contains a large amount of Zircaloy and the containment is relatively small. Because containment leakage cannot be totally excluded, hydrogen can build up in the reactor building, where the atmosphere is normal air. The objective of the work was to investigate, whether hydrogen can form flammable and detonable mixtures in the reactor building, evaluate the possibility of onset of detonation and assess the pressure loads under detonation conditions. The safety concern is, whether the hydrogen in the reactor building can detonate and whether the external detonation can jeopardize the containment integrity. The analysis indicated that the possibility of flame acceleration and deflagration-to-detonation transition (DDT) in the reactor building could not be ruled out in case of a 20 mm 2 leakage from the containment. The detonation analyses indicated that maximum pressure spike of about 7 MPa was observed in the reactor building room selected for the analysis

  17. Earthquake-proof supporting structure in reactor vessel

    International Nuclear Information System (INIS)

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

    1990-01-01

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

  18. Auxiliary water supply device for BWR type reactor

    International Nuclear Information System (INIS)

    Sasagawa, Hiroshi.

    1994-01-01

    In the device of the present invention, a cooling condensation means is disposed to a steam discharge channel of a turbine for driving pumps to directly return condensates to the reactor, so that the temperature of the suppression pool water is not elevated. Namely, the cooling condensation means for discharged steams is disposed to the discharge channel of the turbine. The condensate channel from the cooling condensation means is connected to a sucking side of the turbine driving pump. With such a constitution, when the reactor is isolated from a main steam system, reactor scram is conducted. Although the reactor water level is lowered by the reactor scram, the lowering of the reactor water level is prevented by supplementing cooling water by the turbine driving pump using steams generated in the reactor as a power source. The discharged steams after driving the turbine are cooled and condensated by the cooling condensation means by way of the discharge channel and returned to the reactor again by way of the condensate channel. With such procedures, since the temperature of suppression pool water is not elevated, there is no need to operate other cooling systems. In addition, auxiliary water can be supplied for a long period of time. (I.S.)

  19. Apparatus for locating inspection device in a nuclear reactor vessel

    International Nuclear Information System (INIS)

    1980-01-01

    A method for accurately locating an inspection device with a PWR or BWR pressure vessel uses a plurity of guide members and an internal location element, the exact position of which is known. Used for defining the size, orientation and position of a flow. (U.K.)

  20. 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.

  1. Artificial intelligence applied to fuel management in BWR type reactors; Inteligencia artificial aplicada a la administracion de combustible en reactores BWR

    Energy Technology Data Exchange (ETDEWEB)

    Ortiz S, J.J

    1998-10-01

    In this work two techniques of artificial intelligence, neural networks and genetic algorithms were applied to a practical problem of nuclear fuel management; the determination of the optimal fuel reload for a BWR type reactor. This is an important problem in the design of the operation cycle of the reactor. As a result of the application of these techniques, comparable or even better reloads proposals than those given by expert companies in the subject were obtained. Additionally, two other simpler problems in reactor physics were solved: the determination of the axial power profile and the prediction of the value of some variables of interest at the end of the operation cycle of the reactor. Neural networks and genetic algorithms have been applied to solve many problems of engineering because of their versatility but they have been rarely used in the area of fuel management. The results obtained in this thesis indicates the convenience of undertaking further work on this area and suggest the application of these techniques of artificial intelligence to the solution of other problems in nuclear reactor physics. (Author)

  2. BWR internals life assurance

    International Nuclear Information System (INIS)

    Herrera, M.L.; Stancavage, P.P.

    1988-01-01

    Boiling water reactor (BWR) internal components play an important role in power plant life extension. Many important internals were not designed for easy removal and changes in material properties and local environmental effects due to high radiation makes stress corrosion cracking more likely and more difficult to correct. Over the past several years, operating experience has shown that inspection, monitoring and refurbishment can be accomplished for internal structures with existing technology. In addition, mitigation techniques which address the causes of degradation are available to assure that life extension targets can be met. This paper describes the many considerations and aspects when evaluating life extension for reactor vessel internals

  3. 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)

  4. Stress corrosion cracking of L-grade stainless steels in boiling water reactor (BWR) plants

    International Nuclear Information System (INIS)

    Suzuki, Shunichi; Fukuda, Toshihiko; Yamashita, Hironobu

    2004-01-01

    L-grade stainless steels as 316NG, SUS316L and SUS304L have been used for the BWR reactor internals and re-circulation pipes as SCC resistant materials. However, SCC of the L-grade material components were reported recently in many Japanese BWR plants. The detail investigation of the components showed the fabrication process such as welding, machining and surface finishing strongly affected SCC occurrence. In this paper, research results of SCC of L-grade stainless steels, metallurgical investigation of core shrouds and re-circulation pipings, and features of SCC morphology were introduced. Besides, the structural integrity of components with SCC, countermeasures for SCC and future R and D planning were introduced. (author)

  5. Pressure vessel for nuclear reactors

    International Nuclear Information System (INIS)

    Schulten, R.; Kugeler, K.; Kugeler, M.; Petersen, K.; Decken, C.B. von der.

    1983-01-01

    This construction of a container, which is pressure-relieved by axial-central tensioning cables or reinforcing cables distributed over the circumference, makes a reduction of the wall thickness for the floor and roof, which was previously 2.5 metres by about 40% possible, and thus reduce manufacturing and cost problems. This is achieved by an appreciable increase of the prestressing exerted by the tensioning cables as this is taken up, not by the elasticity of the roof and floor, but instead by an intermediate part of pressure-resisting material. Such a container consists of a vertical cylindrical jacket of, for example, 20 metres diameter and 18 metres height, of a roof and floor of, for example, 1.50 metres thickness each and the intermediate part, which keeps the spacing of floor and roof as a central piece. This intermediate part which is taken through seals through the container can be imagined as a double tube of outside tube diameter of, for example, 4 metres and inside tube diameter of 2 metres with both tubes having thick walls. 4 tensioning cables displaced vertically by 900 run in the cylindrical annulus between the outer and inner tubes which are brought to the required pretension, e.g. 80,000 tonnes by nuts situated on the outside. The inner tube projects through the floor and roof. Its openings act as manholes and for the introduction of pipelines. These can, for example, carry a cooling medium for a reactor core via further ducts into the inside of the container. Container wall, floor and roof and the intermediate part in the form of a double tube are made up of cast steel segments or sectors in several layers. (RW)

  6. 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

  7. Recycling flow rate control device in BWR type reactor

    International Nuclear Information System (INIS)

    Fujiwara, Tadashi; Koda, Yasushi

    1988-01-01

    Purpose: To reduce the recycling pump speed if the pressure variation width and the variation ratio in the nuclear reactor exceed predetermined values, to thereby avoid the shutdown of the plant. Constitution: There has been proposed a method of monitoring the neutron flux increase thereby avoiding unnecessary plant shutdown, but it involves a problems of reactor scram depending on the state of the plant and the set values. In view of the above, in the plant using internal pumps put under the thyristor control and having high response to recycling flow rate, the reactor pressure is monitored and the speed of the internal pump is rapidly reduced when the pressure variation width and variation ratio exceed predetermined values to reduce the reactor power and avoid the plant shutdown. This can reduce the possibility of unnecessary power reduction due to neutron flux noises or the possibility of plant shutdown under low power conditions. Further, since the reactor operation can be continued without stopping the recycling pump, the operation upon recovery can be made rapid. (Horiuchi, T.)

  8. 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.

  9. Identification of process dynamics. Stability monitoring in BWR type reactors

    International Nuclear Information System (INIS)

    Abrahamsson, P.; Hallgren, P.

    1991-06-01

    Identification of process dynamics is used for stability monitoring in nuclear reactors (Boiling Water Reactor). This report treats the problem of estimating a damping factor and a resonance frequency from the neutron flux as measured in the reactor. A new parametric online method for identification is derived and presented, and is shown to meet the requirements of stability monitoring. The technique for estimating the process parameters is based on a recursive lattice filter algorithm. The problem of time varying parameters and offset, as well as offline experiments and signal processing are treated. All parts are implemented in a realtime program, using the language C. In comparison with earlier identifications, the new way of estimating the damping factor is shown to work well. Estimates of both the damping factor and the resonance frequency show a stable and reliable behavior. Future development and improvements are also indicated. (au)

  10. Device for detecting neutron flux in nuclear reactor. [BWR

    Energy Technology Data Exchange (ETDEWEB)

    Bessho, Y; Nishizawa, Y

    1976-07-30

    The object of the invention is to ensure accuracy in the operation of the nuclear reactor by reducing the difference that results between the readings of a Traversing Incore Probe (TIP) and a Local Power Range Monitor (LPRM) when the neutron flux distribution undergoes a change. In an apparatus for detecting neutrons in a nuclear reactor, an LPRM sensor comprising a layer containing a substance capable of nuclear fission, a section filled with argon gas and a collector is constructed so as to surround a TIP within a TIP guide tube at the height of the reactor axis. In this way, the LPRM detects the average value of neutron distribution in the region surrounding the TIP, so that no great difference between the readings of both the sensors is produced even if the neutron flux distribution is changed.

  11. 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

  12. 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

  13. 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)

  14. 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

  15. PWR and BWR light water reactor systems in the USA and their fuel cycle

    International Nuclear Information System (INIS)

    Crawford, W.D.

    1977-01-01

    Light water reactor operating experience in the USA can be considered to date from the choice of the pressurized water reactor (PWR) for use in the naval reactor program and the subsequent construction and operation of the nuclear power plant at Shippingport, Pennsylvania in 1957. The development of the boiling water reactor (BWR) in 1954 and its selection for the plant at Dresden, Illinois in 1959 established this concept as the other major reactor type in the US nuclear power program. The subsequent growth profile is presented, leading to 31 PWR's and 23 BWR's currently in operation as well as to plants in the planning and construction phase. A significant operating record has been accumulated concerning the availability of each of these reactor types as determined by: (1) outage for refueling, (2) component reliability, (3) maintenance requirements, and (4) retrofitting required by government regulation. In addition, the use and performance of BWR's and PWR's in meeting system load requirements is discussed. The growing concern regarding possible terrorist activities and other potential threats has resulted in systems and procedures designed to assure effective safeguards at nuclear power installations. Safeguards measures currently in place are described. Environmental effects of operating plants are subject to both radiological and non-radiological monitoring to verify that results are within the limits established in the licensing process. The operating results achieved and the types of modifications that have been required of operating plants by the Nuclear Regulatory Commission are reviewed. The PWR and BWR Fuel Cycle is examined in terms of: (1) fuel burnup experience and prospects for improvement, (2) the status and outlook for natural uranium resources, (3) enrichment capacity, (4) reprocessing and recycle, and the interrelationships among the latter three factors. High level waste management currently involving on-site storage of spent fuel is discussed

  16. Current status of light water reactor and Hitachi's technical improvements for BWR

    International Nuclear Information System (INIS)

    Miki, Minoru; Ohki, Arahiko.

    1984-01-01

    Gradual technical improvements in Japan over the years has improved the reliability of light water reactors, and has achieved the highest capacity factor level in the world. Commercial operation of Fukushima 2-2 (1,100 MW) of the Tokyo Electric Power Co. was started in February, 1984, as the first standardized BWR base plant, ushering in a new age of domestic light water reactor technology. The ABWR (1,300 MW class) has been developed as Japan's next generation light water reactor, with construction aimed at the latter half of the 1980's. Hitachi's extensive efforts range from key nuclear equipment to various related robots, directed at improving safety, reliability, and the capacity factor, while reducing radiation exposure. This paper presents an outline of Hitachi's participation in the light water reactor's improvement and standardization, and the current status of our role in the international cooperation plan for the ABWR. (author)

  17. 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

  18. 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)

  19. Study on vertical seismic response model of BWR-type reactor building

    International Nuclear Information System (INIS)

    Konno, T.; Motohashi, S.; Izumi, M.; Iizuka, S.

    1993-01-01

    A study on advanced seismic design for LWR has been carried out by the Nuclear Power Engineering Corporation (NUPEC), under the sponsorship of the Ministry of International Trade and Industry (MITI) of Japan. As a part of the study, it has been investigated to construct an accurate analytical model of reactor buildings for a seismic response analysis, which can reasonably represent dynamic characteristics of the building. In Japan, vibration models of reactor buildings for horizontal ground motion have been studied and examined through many simulation analyses for forced vibration tests and earthquake observations of actual buildings. And now it is possible to establish a reliable horizontal vibration model on the basis of multi-lumped mass and spring model. However, vertical vibration models have not been so much studied as horizontal models, due to less observed data for vertical motions. In this paper, the vertical seismic response models of a BWR-type reactor building including soil-structure interaction effect are numerically studied, by comparing the dynamic characteristics of (1) three dimensional finite element model, (2) multi-stick lumped mass model with a flexible base-mat, (3) multi-stick lumped mass model with a rigid base-mat and (4) single-stick lumped mass model. In particular, the BWR-type reactor building has the long span truss roof which is considered to be one of the critical members to vertical excitation. The modelings of the roof trusses are also studied

  20. 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

  1. Structural integrity and management of aging in internal components of BWR reactors; Integridad estructural y manejo del envejecimiento en componentes internos de reactores BWR

    Energy Technology Data Exchange (ETDEWEB)

    Arganis J, C.R. [Instituto Nacional de Investigaciones Nucleares, Km 36.5 Carretera Mexico, Toluca Salazar Edo. de Mexico (Mexico)]. E-mail: craj@nuclear.inin.mx

    2004-07-01

    Presently work the bases to apply structural integrity and the handling of the aging of internal components of the pressure vessel of boiling water reactors of water are revised and is carried out an example of structural integrity in the horizontal welding H4 of the encircling one of the core of a reactor, taking data reported in the literature. It is also revised what is required to carry out the handling program or conduct of the aging (AMP). (Author)

  2. 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

  3. Spent fuel data base: commercial light water reactors. [PWR; BWR

    Energy Technology Data Exchange (ETDEWEB)

    Hauf, M.J.; Kniazewycz, B.G.

    1979-12-01

    As a consequence of this country's non-proliferation policy, the reprocessing of spent nuclear fuel has been delayed indefinitely. This has resulted in spent light water reactor (LWR) fuel being considered as a potential waste form for disposal. Since the Nuclear Regulatory Commission (NRC) is currently developing methodologies for use in the regulation of the management and disposal of high-level and transuranic wastes, a comprehensive data base describing LWR fuel technology must be compiled. This document provides that technology baseline and, as such, will support the development of those evaluation standards and criteria applicable to spent nuclear fuel.

  4. Gas/liquid separator for BWR type reactor

    International Nuclear Information System (INIS)

    Soma, Naoshi; Akimoto, Seiichi; Yokoyama, Iwao.

    1993-01-01

    A two phase gas/liquid flow generated at a heating portion of a nuclear reactor is swirled by inlet vanes. The phase gas/liquid flow uprises as a vortex flow in a vortex cylinder, and a liquid phase of a high density gathers at the outer circumference of the vortex cylinder. The liquid phase gathered at the outer circumference is collected at the inlet of a discharge flow channel which protrude into the vortex cylinder and in a three-step structure, and introduced into a recycling liquid phase passing through the discharge flow channel for liquid phase. There is provided a structure that separated liquid collected at the lowermost state in the inlet of the three-step discharge flow channel inlet descends in the discharge flow channel, then uprises in an uprising flow channel and is introduced into the recycling liquid phase by way of a discharge flow channel exit. The height of the discharge flow channel exit is determined equal to that of a liquid level of the recycling liquid phase during rated operation of the reactor. Accordingly, even in a case where the liquid level in the recycling liquid phase is lowered, the liquid level of the uprising flow channel is kept equal to that during rated operation. (I.N.)

  5. Evaluation of the radial design of fuel cells in an operation cycle of a BWR reactor

    International Nuclear Information System (INIS)

    Gonzalez C, J.; Martin del Campo M, C.

    2003-01-01

    This work is continuation of one previous in the one that the application of the optimization technique called Tabu search to the radial design of fuel cells of boiling water reactors (BWR, Boiling Water Reactor) is presented. The objective function used in the optimization process only include neutron parameters (k-infinite and peak of radial power) considering the cell at infinite media. It was obtained to reduce the cell average enrichment completing the characteristics of reactivity of an original cell. The objective of the present work is to validate the objective function that was used for the radial design of the fuel cell (test cell), analyzing the operation of a one cycle of the reactor in which fuels have been fresh recharged that contain an axial area with the nuclear database of the cell designed instead of the original cell. For it is simulated it with Cm-Presto the cycle 10 of the reactor operation of the Unit 1 of the Nuclear Power station of Laguna Verde (U1-CNLV). For the cycle evaluation its were applied so much the simulation with the Haling strategy, as the simulation of the one cycle with control rod patterns and they were evaluated the energy generation and several power limits and reactivity that are used as design parameters in fuel reloads of BWR reactors. The results at level of an operation cycle of the reactor, show that the objective function used in the optimization and radial design of the cell is adequate and that it can induce to one good use of the fuel. (Author)

  6. 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.)

  7. Analysis CFD for the hydrogen transport in the primary containment of a BWR; Analisis CFD para el transporte de hidrogeno en la contencion primaria de un reactor BWR

    Energy Technology Data Exchange (ETDEWEB)

    Jimenez P, D. A.; Del Valle G, E. [IPN, Escuela Superior de Fisica y Matematicas, Av. IPN s/n, Edificio 9, Col. San Pedro Zacatenco, 07738 Mexico D. F. (Mexico); Gomez T, A. M., E-mail: guerreroazteca_69@hotmail.com [ININ, Departamento de Sistemas Nucleares, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico)

    2014-10-15

    This study presents a qualitative and quantitative comparison among the CFD GASFLOW and OpenFOAM codes which are related with the phenomenon of hydrogen transport and other gases in the primary containment of a Boiling Water Reactor (BWR). GASFLOW is a commercial license code that is well validated and that was developed in Germany for the analysis of the gases transport in containments of nuclear reactors. On the other hand, OpenFOAM is an open source code that offers several evaluation solvers for different types of phenomena; in this case, the solver reacting-Foam is used to analyze the hydrogen transport inside the primary containment of the BWR. The results that offer the solver reacting-Foam of OpenFOAM are evaluated in the hydrogen transport calculation and the results are compared with those of the program of commercial license GASFLOW to see if is viable the use of the open source code in the case of the hydrogen transport in the primary containment of a BWR. Of the obtained results so much quantitative as qualitative some differences were identified between both codes, the differences (with a percentage of maximum error of 4%) in the quantitative results are small and they are considered acceptable for this analysis type, also, these differences are attributed mainly to the used transport models, considering that OpenFOAM uses a homogeneous model and GASFLOW uses a heterogeneous model. (Author)

  8. 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

  9. Method of operating control rods for BWR type reactors

    International Nuclear Information System (INIS)

    Shirakawa, Toshihisa.

    1979-01-01

    Purpose: To eliminate the danger such as fuel element failures due to rapid power increase and form a control rod pattern for obtaining a required power level in a relatively short time. Method: Control rods are disposed so as to vertically enter into and retract from the central region of each four fuel assemblies adjacent to each other respectively. Upon operation of the control rods, every other control rods in the lateral and longitudinal directions among the entire control rods that are inserted completely are extracted completely at the lower flow limit of coolants. Then, the control rods completely inserted are divided into groups inserted deeply and groups inserted less deeply. The less deeply inserted groups are extracted just before the excess of thermal limit value successively in the lower flow limit of the coolants and then the deeply inserted groups are extracted successively till a predetermined power level in the same manner. Therefore, the coolant flow to the reactor core is increased and the power level is raised. (Furukawa, Y.)

  10. Fuel assembly for use in BWR type reactor

    International Nuclear Information System (INIS)

    Inaba, Yuzo.

    1988-01-01

    Purpose: To attain the reduction of neutron irradiation amount to control rods by the improvement in the reactor shutdown margin and the improvement of the control rod worth, by enhancing the arrangement of burnable poisons. Constitution: The number of burnable poison-incorporated fuel rods present in the outer two rows along the sides in adjacent with a control rod among the square lattice arrangement in a fuel assembly is decreased to less than 1/4 for that of total burnable poison-incorporated fuel rods, while the remaining burnable posion-incorporated fuel rods are arranged in the region other than above (that is, those regions not nearer to the control rod). Thus, even if a sufficient number of burnable poison to prolong the controlling effect for the reactivity with the burnable contents as the fuel assembly are disposed, only the burnable poison -incorporated fuel rods by the number less than 1/4 for that of the total burnable poison-incorporated fuel rods are present near the control rod of the fuel assembly. Accordingly, the control rod worth at the initial stage of the burning is increased at both high and normal temperatures. (Kawakami, Y.)

  11. CFD Simulation of rigid venting of the containment of a BWR-5 Mark-II reactor

    International Nuclear Information System (INIS)

    Galindo G, I. F.; Vazquez B, A. K.; Velazquez E, L.; Tijerina S, F.; Tapia M, R.

    2016-09-01

    In conditions of prolonged loss of external energy or a severe accident, venting to the atmosphere is an alternative to prevent overpressure and release of fission products from the primary containment of a nuclear reactor. Due to the importance of flow determination through rigid vents, a computational fluid dynamics (CFD) model is proposed to verify the capacity of rigid vents in the primary containment of a boiling water reactor (BWR) under different operating conditions (pressure, temperature and compositions of the fluids). The model predicts and provides detailed information on variables such as mass flow and velocity of the venting gases. In the proposed model the primary containment gas is vented to the atmosphere via rigid vents (pipes) from the dry and wet pit. Is assumed that the container is pressurized because is in a defined scenario, and at one point the venting is open and the gas released into the atmosphere. The objective is to characterize the flow and validate the CFD model for the overpressure conditions that occur in an accident such as a LOCA, Sbo, etc. The model is implemented with Ansys-Fluent general-purpose CFD software based on the geometry of the venting ducts of the containment of a BWR. The model is developed three-dimensional and resolves at steady state for compressible flow and includes the effects of the turbulence represented by the Reynolds stress model. The CFD results are compared with the values of a one-dimensional and isentropic model for compressible flow. The relative similarity of results leads to the conclusion that the proposed CFD model can help to predict the rigid venting capacity of the containment of a BWR, however more information is required for full validation of the proposed model. (Author)

  12. The probable types, sizes, positions and orientations of the defects which may appear in connection with manufacture of reactor vessels

    International Nuclear Information System (INIS)

    Bergh, S.

    1980-02-01

    An review of welding technology in manufacture of reactor vessels is made. An inventory of principal defects appearing in connection with manual ARC-welding and coated electrodes is presented. Some important welded joints of BWR reactor vessels are scrutinized. Reheating cracks may appear during stress relief annealing beneath the cladding, and this problem is discussed in the third part. The interest is focussed towards the defects which depend on the conditions during the welding. Slag and incomplete fusion might be found. The review can serve for the guidance of nondestructive testing. The defects are estimated to have the size of a few MM with a maximum to approx. 10 MM right across the weld, possibly with exception for the electroslag welds of the OKG-1 reactor vessel. (GBn)

  13. 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

  14. Reliability analysis of reactor pressure vessel intensity

    International Nuclear Information System (INIS)

    Zheng Liangang; Lu Yongbo

    2012-01-01

    This paper performs the reliability analysis of reactor pressure vessel (RPV) with ANSYS. The analysis method include direct Monte Carlo Simulation method, Latin Hypercube Sampling, central composite design and Box-Behnken Matrix design. The RPV integrity reliability under given input condition is proposed. The result shows that the effects on the RPV base material reliability are internal press, allowable basic stress and elasticity modulus of base material in descending order, and the effects on the bolt reliability are allowable basic stress of bolt material, preload of bolt and internal press in descending order. (authors)

  15. Prediction of the stability of BWR reactors during the start-up process; Prediccion de la estabilidad de reactores BWR durante el proceso de arranque

    Energy Technology Data Exchange (ETDEWEB)

    Ruiz E, J.A.; Castillo D, R. [ININ, Km. 36.5 Carretera Mexico-Toluca, 52045 Salazar, Estado de Mexico (Mexico); Blazquez M, J.B. [Centro de Investigaciones Energetics, Medioambientales y Tecnologicas, Av Complutense 22, 28040 Madrid (Spain)

    2004-07-01

    The Boiling Water Reactors (BWR) are susceptible of uncertainties of power when they are operated to low flows of coolant (W) and high powers (P), being presented this situation mainly in the start-up process. The start-up process could be made but sure if the operator knew the value of the stability index Decay reason (Dr) before going up power and therefore to guarantee the stability. The power and the flow are constantly measures, the index Dr could also be considered its value in real time. The index Dr depends on the power, flow and many other values, such as, the distribution of the flow axial and radial neutronic, the temperature of the feeding water, the fraction of holes and other thermohydraulic and nuclear parameters. A simple relationship of Dr is derived leaving of the pattern reduced of March-Leuba, where three independent variables are had that are the power, the flow and a parameter that it contains the rest of the phenomenology, that is to say all the other quantities that affect the value of Dr. This relationship developed work presently and verified its prediction with data of start-up of commercial reactors could be used for the design of a practical procedure practice of start-up, what would support to the operator to prevent this type of events of uncertainty. (Author)

  16. 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)

  17. 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

  18. A neutronic assessment of the new Spherical Cermets Fuel concept for the BWR-PB reactor

    International Nuclear Information System (INIS)

    Benchrif, A.; Chetaine, A.; Amsil, H.; Bounakhla, M.

    2010-01-01

    The tri-structural-isotopic (TRISO) fuel directly cooled by boiling light water is used in the boiling water reactor with pebble-bed coated particles (BWR-PB). At the lower coolant temperature, the TRISO fuel particles demonstrate an unacceptable irradiation swelling in the silicon carbide coating layer during a fuel cycle. So, the objectives of this paper, on the one hand is to evaluate some neutronic parameters of a new fuel concept, Spherical Cermets Fuel (SCF), for a BWR-PB reactor. On the other hand, to assess the fact of SCF fuel concept on the fuel assembly lifetime and the burn-up characteristic. All the parameters as well as Infinite Multiplication Factor, Spectrum Index, Instantaneous Conversion Ratio and Neutron Energy Spectrum was calculated then compared for the TRISO and the SCF fuel concept. It can be seen from the assessment of fuel assembly burn-up characteristics that the normalised neutron spectra of all the assembly's parts pointed out a thermal spectrum for the SCF fuel assembly's parts than the TRISO one. The SCF fuel element increase the assembly life time about 6.1 EFPY corresponding 8000 MWd/t. So, the fuel assembly can be operated for a reasonably long period without outside refuelling. The difference in the assembly lifetime might leads to SCF fuel concept adopted, because the geometry and concept of TRISO fuel particles are wholly different to SCF ones. (author)

  19. 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

  20. 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

  1. 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

  2. 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

  3. 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

  4. 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

  5. Modelling of the dynamics of the vessel and circuits of recirculation of a BWR type nucleo electric as part of the SUN-RAH university simulator

    International Nuclear Information System (INIS)

    Sanchez S, R.A.

    2003-01-01

    In the present project, the development of a model for the dynamics of the process of energy transport generated in the nuclear fuel until the main steam lines of a nucleo electric central with BWR type nuclear reactor, using mathematical models of reduced order is presented. These models present the main characteristics of the reactor vessel and of the recirculation system, defined by the main phenomena that intervene in those physical processes. Likewise, the objective of the general project of the one University student nucleo electric simulator with Boiling Water Reactor (SUN-RAH) for later on to establish the modeling equations for each part of the nuclear reactor as well as of the load pursuit system. Also, its were described the graphic interfaces implemented in an three layers architecture in which the different measuring variables are presented in the monitor. It fits signalize that the advantage presented by the University student nucleo electric simulator is the possibility to carry out changes in the magnitudes of those different variables that intervene in the physical processes made in the one reactor and in the recirculation system in execution time of the same one. Of same way, the creation of a graphic intuitive interface, friendly, and designed with the same technology with the one that the video games are programmed in the present time. Besides all the above mentioned, the pending goals inside of the project are exposed, as well as the developments in construction process or conceptualized to be included in future versions of the simulator. Finally its are thinking about possible scenarios of applications of SUN-RAH, as well as their reaches. (Author)

  6. 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

  7. Assessment of reactor vessel integrity (ARVI)

    Energy Technology Data Exchange (ETDEWEB)

    Sehgal, B.R. [Division of Nuclear Power Safety (NPS), Royal Institute of Technology (KTH), Drottning Kristinas Vaeg 33A, 10044 Stockholm (Sweden)]. E-mail: sehgal@ne.kth.se; Karbojian, A. [Division of Nuclear Power Safety (NPS), Royal Institute of Technology (KTH), Drottning Kristinas Vaeg 33A, 10044 Stockholm (Sweden); Giri, A. [Division of Nuclear Power Safety (NPS), Royal Institute of Technology (KTH), Drottning Kristinas Vaeg 33A, 10044 Stockholm (Sweden); Kymaelaeinen, O. [FortumEngNP (Finland); Bonnet, J.M. [CEA (France); Ikkonen, K. [Division of Nuclear Power Safety (NPS), Royal Institute of Technology (KTH), Drottning Kristinas Vaeg 33A, 10044 Stockholm (Sweden); Sairanen, R. [VTT (Finland); Bhandari, S. [FRAMATOME (France); Buerger, M. [USTUTT (Germany); Dienstbier, J. [NRI Rez (Czech Republic); Techy, Z. [VEIKI (Hungary); Theofanous, T. [UCSB (United States)

    2005-02-01

    The assessment of reactor vessel integrity (ARVI) project involved a total of nine organizations from Europe and USA. The work consisted of experiments and analysis development. The modeling activities in the area of structural analyses were focused on the support of EC-FOREVER experiments as well as on the exploitation of the data obtained from those experiments for modeling of creep deformation and the validation of the industry structural codes. Work was also performed for extension of melt natural convection analyses to consideration of stratification, and mixing (in the CFD codes). Other modeling activities were for (1) gap cooling CHF and (2) developing simple models for system code. Finally, the methodology and data was applied for the design of IVMR severe accident management scheme for VVER-440/213 plants. The work was broken up into five packages. They were divided into tasks, which were performed by different partners. The major experimental project continued was EC-FOREVER in which data was obtained on in-vessel melt pool coolability. In previous EC-FOREVER experiments data was obtained on melt pool natural convection and lower head creep failure and rupture. Those results obtained were related to the following issues: (1) multiaxial creep laws for different vessel steels (2) effects of penetrations, and (3) mode and location of lower head failure. The two EC-FOREVER tests reported here are related to (a) the effectiveness of gap cooling and (b) water ingression for in vessel melt coolability. Two other experimental projects were also conducted. One was the COPO experiments, which was concerned with the effects of stratification and metal layer on the thermal loads on the lower head wall during melt pool convection. The second experimental project was conducted at ULPU facility, which provided data and correlations of CHF due to the external cooling of the lower head.

  8. Residual stress analysis in reactor pressure vessel attachments

    International Nuclear Information System (INIS)

    Dexter, R.J.; Pont, D.

    1991-08-01

    Residual stresses in cladding and welded attachments could contribute to the problem of stress-corrosion cracking in boiling-water reactors (BWR). As part of a larger program aimed at quantifying residual stress in BWR components, models that would be applicable for predicting residual stress in BWR components are reviewed and documented. The review includes simple methods of estimating residual stresses as well as advanced finite-element software. In general, simple methods are capable of predicting peak magnitudes of residual stresses but are incapable of adequately characterizing the distribution of residual stresses. Ten groups of researchers using finite-element software are reviewed in detail. For each group, the assumptions of the model, possible simplifications, material property data, and specific applications are discussed. The most accurate results are obtained when a metallurgical simulation is performed, transformation plasticity effects are included, and the heating and cooling parts of the welding thermal cycle are simulated. Two models are identified which can provide these features. The present state of these models and the material property data available in the literature are adequate to quantify residual stress in BWR components

  9. A practical methodology of radiological protection for the reduction of hot particles in BWR type reactors

    International Nuclear Information System (INIS)

    Alvarez G, G.

    1991-01-01

    The purpose of this work, in general form, is to describe a practical method for reduction of hot particles generated as consequence of the operational activities of BWR nuclear reactors. This methodology provides a description of the localizations and/or probable activities of finding particles highly radioactive denominated hot particles. For this purpose it was developed a strategy based on the decontamination lineaments, as well as the manipulation, gathering, registration, contention, documentation, control and final disposition of the hot particles. In addition, some recommendations are reiterated and alternative, in order to gathering the hot particles in a dynamic way given to the activities of the personal occupationally exposed in highly radioactive areas. The structure of the methodology of hot particles is supported in the radiological controls based on the Code of Federal Regulation 10 CFR 20 as well as the applicable regulatory documents. It provides an idea based on administrative controls of radiological protection, in order to suggesting the responsibilities and necessary directing for the control of the hot particles required in nuclear plants of the BWR type. (author)

  10. Reactor pressure vessel stud management automation strategies

    International Nuclear Information System (INIS)

    Biach, W.L.; Hill, R.; Hung, K.

    1992-01-01

    The adoption of hydraulic tensioner technology as the standard for bolting and unbolting the reactor pressure vessel (RPV) head 35 yr ago represented an incredible commitment to new technology, but the existing technology was so primitive as to be clearly unacceptable. Today, a variety of approaches for improvement make the decision more difficult. Automation in existing installations must meet complex physical, logistic, and financial parameters while addressing the demands of reduced exposure, reduced critical path, and extended plant life. There are two generic approaches to providing automated RPV stud engagement and disengagement: the multiple stud tensioner and automated individual tools. A variation of the latter would include the handling system. Each has its benefits and liabilities

  11. Seal analysis technology for reactor pressure vessel

    International Nuclear Information System (INIS)

    Zheng Liangang; Zhang Liping; Yang Yu; Zang Fenggang

    2009-01-01

    There is the coolant with radiation, high temperature and high pressure in the reactor pressure vessel (RPV). It is closely correlated to RPV sealing capability whether the whole nuclear system work well or not. The aim of this paper is to study the seal analysis method and technology, such as the pre-tensioning of the bolt, elastoplastic contact and coupled technology of thermal and structure. The 3 D elastoplastic seal analysis method really and generally consider the loads and model the contact problem with friction between the contact plates. This method is easier than the specialized seal program and used widely. And it is more really than the 2 D seal analysis method. This 3 D elastoplastic seal analysis method has been successfully used in the design and analysis of RPV. (authors)

  12. Method of estimating thermal power distribution of core of BWR type reactor

    International Nuclear Information System (INIS)

    Sekimizu, Koichi

    1982-01-01

    Purpose: To accurately and rapidly predict the thermal power of the core of a BWR they reactor at load follow-up operating time. Method: A parameter value corrected from a correction coefficient deciding unit and a xenon density distribution value predicted and calculated from a xenon density distributor are inputted to a thermal power distribution predicting devise, the status amount such as coolant flow rate or the like predetermined at this and next high power operating times is substituted for physical model to predict and calculate the thermal power distribution. The status amount of a nuclear reactor at the time of operating in previous high power corresponding to the next high power operation to be predicted is read from the status amount of the reactor stored in time series manner is a reactor core status memory, and the physical model used in the prediction and calculation of the thermal power distribution at the time of next high power operation is corrected. (Sikiya, K.)

  13. An advanced frequency-domain code for boiling water reactor (BWR) stability analysis and design

    International Nuclear Information System (INIS)

    Behrooz, A.

    2008-01-01

    for a large reactor could reach 20 gigabytes) that it is not possible to load into RAM memory of an operating system with 32 bit architecture. A special procedure has been developed within the MATLAB environment to remove this memory limitation, and to invert such large matrices and finally obtain the reactor transfer functions that enable the study of system stability. Various applications of the present frequency-domain code to a typical BWR fuel assembly, a BWR core, and to a chemical reactor showed a good agreement with reference results. (author)

  14. Reactor pressure vessel structural integrity research

    International Nuclear Information System (INIS)

    Pennell, W.E.; Corwin, W.R.

    1994-01-01

    Development continues on the technology used to assess the safety of irradiation-embrittled nuclear reactor pressure vessels (RPVs) containing flaws. Fracture mechanics tests on RPV steel, coupled with detailed elastic-plastic finite-element analyses of the crack-tip stress fields, have shown that (1) constraint relaxation at the crack tip of shallow surface flaws results in increased data scatter but no increase in the lower-bound fracture toughness, (2) the nil ductility temperature (NDT) performs better than the reference temperature for nil ductility transition (RT NDT ) as a normalizing parameter for shallow-flaw fracture toughness data, (3) biaxial loading can reduce the shallow-flaw fracture toughness, (4) stress-based dual-parameter fracture toughness correlations cannot predict the effect of biaxial loading on shallow-flaw fracture toughness because in-plane stresses at the crack tip are not influenced by biaxial loading, and (5) an implicit strain-based dual-parameter fracture toughness correlation can predict the effect of biaxial loading on shallow-flaw fracture toughness. Experimental irradiation investigations have shown that (1) the irradiation-induced shift in Charpy V-notch vs temperature behavior may not be adequate to conservatively assess fracture toughness shifts due to embrittlement, and (2) the wide global variations of initial chemistry and fracture properties of a nominally uniform material within a pressure vessel may confound accurate integrity assessments that require baseline properties

  15. 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)

  16. BWR [boiling water reactor] core criticality versus water level during an ATWS [anticipated transient without scram] event

    International Nuclear Information System (INIS)

    Sehgal, B.R.; Peng, C.M.; Maly, J.

    1988-01-01

    The BWR [boiling water reactor] emergency procedures guidelines recommend management of core water level to reduce the power generated during an anticipated transient without scram (ATWS) event. BWR power level variation has traditionally been calculated in the system codes using a 1-D [one-dimensional] 2-group neutron kinetics model to determine criticality. This methodology used also for calculating criticality of the partially covered BWR cores has, however, never been validated against data. In this paper, the power level versus water level issues in an ATWS severe accident are introduced and the accuracy of the traditional methodology is investigated by comparing with measured data. It is found that the 1-D 2-group treatment is not adequate for accurate predictions of criticality and therefore the system power level for the water level variations that may be encountered in a prototypical ATWS severe accident. It is believed that the current predictions for power level may be too high

  17. 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)

  18. 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

  19. 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.

  20. Structural integrity and management of aging in internal components of BWR reactors

    International Nuclear Information System (INIS)

    Arganis J, C.R.

    2004-01-01

    Presently work the bases to apply structural integrity and the handling of the aging of internal components of the pressure vessel of boiling water reactors of water are revised and is carried out an example of structural integrity in the horizontal welding H4 of the encircling one of the core of a reactor, taking data reported in the literature. It is also revised what is required to carry out the handling program or conduct of the aging (AMP). (Author)

  1. Analysis of multidimensional and countercurrent effects in a BWR loss-of-coolant accident

    International Nuclear Information System (INIS)

    Shiralkar, B.S.; Dix, G.E.; Alamgir, M.

    1991-01-01

    The presence of parallel enclosed channels in a boiling water reactor (BWR) provides opportunities for multiple flow regimes in cocurrent and countercurrent flow under loss-of-coolant accident (LOCA) conditions. To address and understand these phenomena, an integrated experimental and analytical study has been conducted. The primary experimental facility was the steam sector test facility (SSFT), which simulated a full scale 30deg sector of a BWR/6 reactor vessel. Both steady-state separate effects tests an integral transients with vessel vlowdown and refill were performed. The presence of multidimensional and parallel-channel effects was found to be very beneficial to BWR LOCA performance. The best estimate TRAC-BWR computer code was extended as part of this study by incorporation of a phenomenological upper plenum mixing model. TRAC-BWR was applied to the analysis of these full scale experiments. Excellent predictions of phenomena and experimental trends were achieved. (orig.)

  2. Development of a methodology of analysis of instabilities in BWR reactors; Desarrollo de una metodologia de analisis de inestabilidades en reactores PWR

    Energy Technology Data Exchange (ETDEWEB)

    Garcia-Fenoll, M.; Abarca, A.; Barrachina, T.; Miro, R.; Verdu, G.

    2012-07-01

    This paper presents a methodology of analysis of the reactors instabilities of BWR type. This methodology covers of modal analysis of the point operation techniques of signal analysis and simulation of transients, through 3D Coupled RELAP5/PARCSv2.7 code.

  3. Assessment of two BWR accident management strategies

    International Nuclear Information System (INIS)

    Hodge, S.A.; Petek, M.

    1991-01-01

    Candidate mitigative strategies for management of in-vessel events during the late phase (after core degradation has occurred) of postulated BWR severe accidents were considered at Oak Ridge National Laboratory (ORNL) during 1990. The identification of new strategies was subject to the constraint that they should, to the maximum extent possible, make use of the existing equipment and water resources of the BWR facilities and not require major equipment modifications or additions. As a result of this effort, two of these candidate strategies were recommended for additional assessment. The first is a strategy for containment flooding to maintain the core and structural debris within the reactor vessel in the event that vessel injection cannot be restored to terminate a severe accident sequence. The second strategy pertains to the opposite case, for which vessel injection would be restored after control blade melting had begun; its purpose is to provide an injection source of borated water at the concentration necessary to preclude criticality upon recovering a damaged BWR core. Assessments of these two strategies have been performed during 1991 under the auspices of the Detailed Assessment of BWR In-Vessel Strategies Program. This paper provides a discussion of the motivation for and purpose of these strategies and the potential for their success. 33 refs., 9 figs

  4. Environmentally-Assisted Cracking of Low-Alloy Reactor Pressure Vessel Steels under Boiling Water Reactor Conditions

    Energy Technology Data Exchange (ETDEWEB)

    Seifert, H.P.; Ritter, S

    2002-02-01

    The present report summarizes the experimental work performed by PSI on the environmentally-assisted cracking (EAC) of low-alloy steels (LAS) in the frame of the RIKORR-project during the period from January 2000 to August 2001. Within this project, the EAC crack growth behaviour of different low-alloy reactor pressure vessel (RPV) steels, weld filler and weld heat-affected zone materials is investigated under simulated transient and steady-state BWR/NWC power operation conditions. The EAC crack growth behaviour of different low-alloy RPV steels was characterized by slow rising load (SRL) / low-frequency corrosion fatigue (LFCF) and constant load tests with pre-cracked fracture mechanics specimens in oxygenated high-temperature water at temperatures of either 288, 250, 200 or 150 C. These tests revealed the following important interim results: Under low-flow and highly oxidizing (ECP >= 100 mV SHE) conditions, the ASME XI 'wet' reference fatigue crack growth curve could be significantly exceeded by cyclic fatigue loading at low frequencies (<0.001 Hz), at high and low load-ratios R, and by ripple loading near to DKth fatigue thresholds. The BWR VIP 60 SCC disposition lines may be significantly or slightly exceeded (even in steels with a low sulphur content) in the case of small load fluctuations at high load ratios (ripple loading) or at intermediate temperatures (200 -250 C) in RPV materials, which show a distinct susceptibility to dynamic strain ageing (DSA). (author)

  5. 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)

  6. Comparative analysis of results between CASMO, MCNP and Serpent for a suite of Benchmark problems on BWR reactors; Analisis comparativo de resultados entre CASMO, MCNP y SERPENT para una suite de problemas Benchmark en reactores BWR

    Energy Technology Data Exchange (ETDEWEB)

    Xolocostli M, J. V.; Vargas E, S.; Gomez T, A. M. [ININ, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico); Reyes F, M. del C.; Del Valle G, E., E-mail: vicente.xolocostli@inin.gob.mx [IPN, Escuela Superior de Fisica y Matematicas, UP - Adolfo Lopez Mateos, Edif. 9, 07738 Mexico D. F. (Mexico)

    2014-10-15

    In this paper a comparison is made in analyzing the suite of Benchmark problems for reactors type BWR between CASMO-4, MCNP6 and Serpent code. The Benchmark problem consists of two different geometries: a fuel cell of a pin and assembly type BWR. To facilitate the study of reactors physics in the fuel pin their nuclear characteristics are provided to detail, such as burnt dependence, the reactivity of selected nuclide, etc. With respect to the fuel assembly, the presented results are regarding to infinite multiplication factor for burning different steps and different vacuum conditions. Making the analysis of this set of Benchmark problems provides comprehensive test problems for the next fuels generation of BWR reactors with high extended burned. It is important to note that when making this comparison the purpose is to validate the methodologies used in modeling for different operating conditions, if the case is of other BWR assembly. The results will be within a range with some uncertainty, considering that does not depend on code that is used. Escuela Superior de Fisica y Matematicas of Instituto Politecnico Nacional (IPN (Mexico) has accumulated some experience in using Serpent, due to the potential of this code over other commercial codes such as CASMO and MCNP. The obtained results for the infinite multiplication factor are encouraging and motivate the studies to continue with the generation of the X S of a core to a next step a respective nuclear data library is constructed and this can be used by codes developed as part of the development project of the Mexican Analysis Platform of Nuclear Reactors AZTLAN. (Author)

  7. Uncertainty Characterization of Reactor Vessel Fracture Toughness

    International Nuclear Information System (INIS)

    Li, Fei; Modarres, Mohammad

    2002-01-01

    To perform fracture mechanics analysis of reactor vessel, fracture toughness (K Ic ) at various temperatures would be necessary. In a best estimate approach, K Ic uncertainties resulting from both lack of sufficient knowledge and randomness in some of the variables of K Ic must be characterized. Although it may be argued that there is only one type of uncertainty, which is lack of perfect knowledge about the subject under study, as a matter of practice K Ic uncertainties can be divided into two types: aleatory and epistemic. Aleatory uncertainty is related to uncertainty that is very difficult to reduce, if not impossible; epistemic uncertainty, on the other hand, can be practically reduced. Distinction between aleatory and epistemic uncertainties facilitates decision-making under uncertainty and allows for proper propagation of uncertainties in the computation process. Typically, epistemic uncertainties representing, for example, parameters of a model are sampled (to generate a 'snapshot', single-value of the parameters), but the totality of aleatory uncertainties is carried through the calculation as available. In this paper a description of an approach to account for these two types of uncertainties associated with K Ic has been provided. (authors)

  8. Inductive testing of reactor pressure vessels

    International Nuclear Information System (INIS)

    Bergh, H.

    1987-01-01

    In Service Inspection of Reactor Pressure Vessels is mostly done with ultrasonics. Using special 2 crystal-probes good detectability is achieved for near surface defects. The problem is to detect closely spaced cracks, to decide if the defects are surface braking and, if not, to decide the remaining ligament. The purpose of this study is to investigate to what extent Eddy Current can solve these problems. Detecting surfacebreaking cracks and fields of cracks can be done using conventional Eddy Current techniques. Mapping of closely spaced cracks requires a small probe and a high frequency. Measurement of depths a larger probe, a lower frequency and knowledge of the crackfield since 2 closely spaced shallow cracks might be mistaken for one deep crack. Depths of singel cracks can be measured down to 7-8 mm. In closely spaced crackfields the depths can not be measured. The measurement is mostly based on amplitude. For not surface breaking defects the problem is to decide the ligament, i.e. the distance between surface and cracktip. To achieve good penetration a large probe, low frequency and high energy or pulsed energy is used. Ligament up to 4 mm can be measured with good accuracy. The measurements is mostly based on phase. Noise, which originates from rough surface, varied material structure and lift off, can be reduced using multi frequency mix, probe design and scanning pattern. (author)

  9. Inspection device for external examination of pressure vessels, preferably for ultrasonic testing of reactor vessels

    International Nuclear Information System (INIS)

    Figlhuber, D.; Gallwas, J.; Weber, R.; Weber, J.

    1978-01-01

    The inspection device is placed in the annular gap between pressure vessel and biological shield of the BWR. In the annulus there is arranged at least one longitudinal rail which has got vertical guideways. Along it there can be moved on testing paths a manipulator with the ultrasonic search unit. The manipulator drive is outside of the inspection annulus. It is coupled to the manipulator by means of a tension member being guided over a reversing unit mounted at the upper end of the longitudinal rail. As a tension member there may be used a drag chain; the drive and the reversing unit are provided with corresponding chain wheels. (DG) [de

  10. PWR and BWR light water reactor systems in the USA and their fuel cycle

    International Nuclear Information System (INIS)

    Crawford, W.D.

    1977-01-01

    Light water reactor operating experience in the USA can be considered to date from the choice of the PWR for use in the naval reactor programme and the subsequent construction and operation of the nuclear power plant at Shippingport in 1957. The development of the BWR in 1954 and its selection for the plant at Dresden in 1959 established this concept as the other major reactor type in the US nuclear power programme. The subsequent growth profile is presented. A significant operating record has been accumulated concerning the availability of each of these reactor types. In addition, the use and performance of BWRs and PWRs in meeting system load requirements is discussed. The growing concern regarding possible terrorist activities and other potential threats has resulted in systems and procedures designed to ensure effective safeguards at nuclear power installations; current measures are described. Environmental effects of operating plants are subject to both radiological and non-radiological monitoring. The operating results achieved and the types of modifications that have been required of operating plants by the Nuclear Regulatory Commission are reviewed. Both fuel cycles are examined in terms of: fuel burnup experience and prospects for improvement; natural uranium resources; enrichment capacity; reprocessing and recycle; and the interrelationships among the latter three factors. High-level waste management currently involving on-site storage of spent fuel is discussed in terms of available capacity and plans for expansion. The US electric utility industry viewpoint regarding an ultimate programme for waste management is outlined. Finally, the current economics and future cost trends of nuclear power plants are evaluated. (author)

  11. Prediction of the stability of BWR reactors during the start-up process

    International Nuclear Information System (INIS)

    Ruiz E, J.A.; Castillo D, R.; Blazquez M, J.B.

    2004-01-01

    The Boiling Water Reactors (BWR) are susceptible of uncertainties of power when they are operated to low flows of coolant (W) and high powers (P), being presented this situation mainly in the start-up process. The start-up process could be made but sure if the operator knew the value of the stability index Decay reason (Dr) before going up power and therefore to guarantee the stability. The power and the flow are constantly measures, the index Dr could also be considered its value in real time. The index Dr depends on the power, flow and many other values, such as, the distribution of the flow axial and radial neutronic, the temperature of the feeding water, the fraction of holes and other thermohydraulic and nuclear parameters. A simple relationship of Dr is derived leaving of the pattern reduced of March-Leuba, where three independent variables are had that are the power, the flow and a parameter that it contains the rest of the phenomenology, that is to say all the other quantities that affect the value of Dr. This relationship developed work presently and verified its prediction with data of start-up of commercial reactors could be used for the design of a practical procedure practice of start-up, what would support to the operator to prevent this type of events of uncertainty. (Author)

  12. Comparative analysis of results between CASMO, MCNP and Serpent for a suite of Benchmark problems on BWR reactors

    International Nuclear Information System (INIS)

    Xolocostli M, J. V.; Vargas E, S.; Gomez T, A. M.; Reyes F, M. del C.; Del Valle G, E.

    2014-10-01

    In this paper a comparison is made in analyzing the suite of Benchmark problems for reactors type BWR between CASMO-4, MCNP6 and Serpent code. The Benchmark problem consists of two different geometries: a fuel cell of a pin and assembly type BWR. To facilitate the study of reactors physics in the fuel pin their nuclear characteristics are provided to detail, such as burnt dependence, the reactivity of selected nuclide, etc. With respect to the fuel assembly, the presented results are regarding to infinite multiplication factor for burning different steps and different vacuum conditions. Making the analysis of this set of Benchmark problems provides comprehensive test problems for the next fuels generation of BWR reactors with high extended burned. It is important to note that when making this comparison the purpose is to validate the methodologies used in modeling for different operating conditions, if the case is of other BWR assembly. The results will be within a range with some uncertainty, considering that does not depend on code that is used. Escuela Superior de Fisica y Matematicas of Instituto Politecnico Nacional (IPN (Mexico) has accumulated some experience in using Serpent, due to the potential of this code over other commercial codes such as CASMO and MCNP. The obtained results for the infinite multiplication factor are encouraging and motivate the studies to continue with the generation of the X S of a core to a next step a respective nuclear data library is constructed and this can be used by codes developed as part of the development project of the Mexican Analysis Platform of Nuclear Reactors AZTLAN. (Author)

  13. Inspection and repair of reactor pressure vessel (RPV) internals

    International Nuclear Information System (INIS)

    Bohmann, W.; Poetz, F.; Nicolai, M.

    1996-01-01

    The past 10 years of operation of light water reactors were characterized by intensive inspection- and repair work on vital components. For boiling water reactors (BWR) it was typical to totally replace the piping system and for pressurized water reactors (PWR) it was the step to complete steam generator (SG) replacement - besides the development of increasingly diligent inspection and repair methods for SG tubes. It can be expected that in the 10 years to come the development of inspection- and repair methods will be aimed mainly at the core internals of BWR's as well as PWR's. Our prediction is that before the end of this decade a first complete replacement of these components will be performed. Already to date a broad range of techniques are available which enable the utilities to carry out inspections and repair of components of core internals in a relatively short time and acceptable expenses. Using examples such as Fuel Alignment Pin Inspection and Replacement, Baffle Former Bolt Inspection and Replacement, Core Barrel Former Bolt Inspection which are typical for PWR's we will in the following describe the existing methods, their development and - last but not least - their successful utilization. What is going to happen in the future? Ageing of the operating plants will continue, thus requesting the plant operators as well as the service companies to work on advanced technologies to fulfill the needs of the industry. (author)

  14. Solution of a benchmark set problems for BWR and PWR reactors with UO2 and MOX fuels using CASMO-4

    International Nuclear Information System (INIS)

    Martinez F, M.A.; Valle G, E. del; Alonso V, G.

    2007-01-01

    In this work some of the results for a group of benchmark problems of light water reactors that allow to study the physics of the fuels of these reactors are presented. These benchmark problems were proposed by Akio Yamamoto and collaborators in 2002 and they include two fuel types; uranium dioxide (UO 2 ) and mixed oxides (MOX). The range of problems that its cover embraces three different configurations: unitary cell for a fuel bar, fuel assemble of PWR and fuel assemble of BWR what allows to carry out an understanding analysis of the problems related with the fuel performance of new generation in light water reactors with high burnt. Also these benchmark problems help to understand the fuel administration in core of a BWR like of a PWR. The calculations were carried out with CMS (of their initials in English Core Management Software), particularly with CASMO-4 that is a code designed to carry out analysis of fuels burnt of fuel bars cells as well as fuel assemblies as much for PWR as for BWR and that it is part in turn of the CMS code. (Author)

  15. Age-related degradation of boiling water reactor vessel internals

    International Nuclear Information System (INIS)

    Ware, A.G.; Shah, V.N.

    1992-01-01

    Researchers at the Idaho National Engineering Laboratory performed an assessment of the aging of the reactor internals in boiling water reactors (BWRs), and identified the unresolved technical issues related to the degradation of these components. The overall life-limiting mechanism is intergranular stress corrosion cracking (IGSCC). Irradiation-assisted stress corrosion cracking, fatigue, and thermal aging embrittlement are other potential degradation mechanisms. Several failures in BWR internals have been caused by a combination of factors such as environment, high residual or preload stresses, and flow-induced vibration. The ASME Code Section XI in-service inspection requirements are insufficient for detecting aging-related degradation at many locations in reactor internals. Many of the potential locations for IGSCC or fatigue are not accessible for inspection. (orig.)

  16. 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

  17. 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

  18. 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)

  19. 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

  20. 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

  1. 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

  2. Detection and characterization of flaws in segments of light water reactor pressure vessels

    International Nuclear Information System (INIS)

    Cook, K.V.; Cunningham, R.A. Jr.; McClung, R.W.

    1988-01-01

    Studies have been conducted to determine flaw density in segments cut from light water reactor )LWR) pressure vessels as part of the Oak Ridge National Laboratory's Heavy-Section Steel Technology (H SST) Program. Segments from the Hope Creek Unit 2 vessel and the Pilgrim Unit 2 Vessel were purchased from salvage dealers. Hope Creek was a boiling water reactor (BWR) design and Pilgrim was a pressurized water reactor (PWR) design. Neither were ever placed in service. Objectives were to evaluate these LWR segments for flaws with ultrasonic and liquid penetrant techniques. Both objectives were successfully completed. One significant indication was detected in a Hope Creek seam weld by ultrasonic techniques and characterized by further analyses terminating with destructive correlation. This indication [with a through-wall dimension of ∼6 mm (∼0.24 in.)] was detected in only 3 m (10 ft) of weldment and offers extremely limited data when compared to the extent of welding even in a single pressure vessel. However, the detection and confirmation of the flaw in the arbitrarily selected sections implies the Marshall report estimates (and others) are nonconservative for such small flaws. No significant indications were detected in the Pilgrim material by ultrasonic techniques. Unfortunately, the Pilgrim segments contained relatively little weldment; thus, we limited our ultrasonic examinations to the cladding and subcladding regions. Fluorescent liquid penetrant inspection of the cladding surfaces for both LWR segments detected no significant indications [i.e., for a total of approximately 6.8 m 2 (72 ft 2 ) of cladding surface]. (author)

  3. 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

  4. 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)

  5. Synergistic failure of BWR internals

    International Nuclear Information System (INIS)

    Ware, A. G.; Chang, T.Y.

    1999-01-01

    Boiling Water Reactor (BWR) core shrouds and other reactor internals important to safety are experiencing intergranular stress corrosion cracking (IGSCC). The United States Nuclear Regulatory Commission has followed the problem, and as part of its investigations, contracted with the Idaho National Engineering and Environmental Laboratory to conduct a risk assessment. The overall project objective is to assess the potential consequences and risks associated with the failure of IGSCC-susceptible BWR vessel internals, with specific consideration given to potential cascading and common mode effects. An initial phase has been completed in which background material was gathered and evaluated, and potential accident sequences were identified. A second phase is underway to perform a simplified, quantitative probabilistic risk assessment on a representative high-power BWR/4. Results of the initial study conducted on the jet pumps show that any cascading failures would not result in a significant increase in the core damage frequency. The methodology is currently being extended to other major reactor internals components

  6. 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.)

  7. An overview of reactor vessel internals segmentation for nuclear plant decommissioning

    International Nuclear Information System (INIS)

    Litka, T.J.

    1994-01-01

    Several nuclear plants have undergone reactor vessel (RV) internals segmentation as part of or in preparation for decommissioning the plant. In addition, several other nuclear facilities are planning for similar work efforts. The primary technology used for segmentation of RV internals, whether in-air or underwater is Plasma Arc Cutting (PAC). Metal Disintegration Machining (MDM) is also used for difficult to make cuts. PAC and MDM are deployed by various means including Long Handled Tools (LHTs), fixtures, tracks, and multi-axis manipulators. These enable remote cutting due to the radiation and/or underwater environment. A Boiling Water Reactor (BWR), a Pressurized Water Reactor (PWR), and a High Temperature Gas Reactor (HTGR) have had their internals removed and segmented using PAC and MDM. The cutting technology used for each component, location of cut, cut geometry and environment had to be determined well before the actual cutting operations. This allowed for the design, fabrication, and testing of the delivery systems. The technologies, selection process, and methodology for RV internals segmentation will be discussed in this paper

  8. 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

  9. Analysis of multi-dimensional and countercurrent effects in a BWR loss-of-coolant accident

    International Nuclear Information System (INIS)

    Shiralkar, B.S.; Dix, G.E.; Alamgir, M.

    1989-01-01

    The presence of parallel enclosed channels in a BWR provides opportunities for multiple flow regimes in co-current and countercurrent flow under Loss-of-Coolant Accident (LOCA) conditions. To address and understand these phenomena, an integrated experimental and analytical study has been conducted. The primary experimental facility was the Steam Sector Test Facility (SSTF) which simulated a full scale 30deg sector of a BWR/6 reactor vessel. Both steady-state separate effects tests and integral transients with vessel blowdown and refill were performed. The present of multi-dimensional and parallel channel effects was found to be very beneficial to BWR LOCA performance. The best estimate TRAC-BWR computer code was extended as part of this study by incorporation of a phenomenological upper plenum mixing model. TRAC-BWR was applied to the analysis of these full scale experiments. Excellent predictions of phenomena and experimental trends were achieved. (orig.)

  10. The BWR Stability Issue

    International Nuclear Information System (INIS)

    D'Auria, F.

    2008-01-01

    The purpose of this paper is to supply general information about Boiling Water Reactor (BWR) stability. The main concerned topics are: phenomenological aspects, experimental database, modelling features and capabilities, numerical models, three-dimensional modelling, BWR system performance during stability, stability monitoring and licensing aspects.

  11. 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

  12. 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

  13. 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

  14. 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

  15. 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)

  16. Radionuclide buildup in BWR [boiling water reactor] reactor coolant recirculation piping

    International Nuclear Information System (INIS)

    Duce, S.W.; Marley, A.W.; Freeman, A.L.

    1989-12-01

    Since the spring of 1985, thermoluminescent dosimeter, dose rate, and gamma spectral data have been acquired on the contamination of boiling water reactor primary coolant recirculation systems as part of a Nuclear Regulatory Commission funded study. Data have been gathered for twelve facilities by taking direct measurements and/or obtaining plant and vendor data. The project titled, ''Effectiveness and Safety Aspects of Selected Decontamination Processes'' (October 1983) initially reviewed the application of chemical decontamination processes on primary coolant recirculation system piping. Recontamination of the system following pipe replacement or chemical decontamination was studied as a second thrust of this program. During the course of this study, recontamination measurements were made at eight different commercial boiling water reactors. At four of the reactors the primary coolant recirculation system piping was chemically decontaminated. At the other four the piping was replaced. Vendor data were obtained from two boiling water reactors that had replaced the primary coolant recirculation system piping. Contamination measurements were made at two newly operating boiling water reactors. This report discusses the results of these measurements as they apply to contamination and recontamination of boiling water reactor recirculation piping. 16 refs., 29 figs., 9 tabs

  17. 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

  18. Determination of stresses caused by fluctuation of acoustic load in the steam dryers of a BWR; Determinacion de esfuerzos originados por fluctuacion de carga acustica en los secadores de vapor de un reactor BWR

    Energy Technology Data Exchange (ETDEWEB)

    Centeno P, J.; Quezada G, S.; Prieto G, A.; Vazquez R, A.; Espinosa P, G. [Universidad Autonoma Metropolitana, Unidad Iztapalapa, Av. San Rafael Atlixco No. 186, Col. Vicentina, 09340 Mexico D. F. (Mexico); Nunez C, A., E-mail: javcuami26@hotmail.com [Comision Nacional de Seguridad Nuclear y Salvaguardias, Dr. Jose Ma. Barragan No. 779, Col. Narvarte, 03020 Mexico D. F. (Mexico)

    2014-10-15

    The extended power up-rate (EPU) in a nuclear power plant cause various problems in BWR components also in the steam system. This due to increased steam flow generated in the reactor and is conveyed to the turbine by the four main steam lines (MSL). One of the most serious problems is the generation of acoustic pressure loads in the metal structure of the steam dryer which eventually leads to fatigue failure and even the appearance of cracks, and in turn it causes loose parts that are entrained by the steam and transported in the MSL. This problem is due to the fluctuation of load acoustics caused by the union of the safety or relief valves (SRV) with the MSL, spreading through these to reach the reactor pressure vessel (RPV) where the effect of resonance of the acoustic wave is amplified and impacts directly in the supporting structure of the steam dryer, skirt and the panels where the mixture liquid-steam is dried, by centrifugation effect and runoff of liquid water. Efforts in the steam dryer operating conditions of EPU for two cases will be analyzed in this work, the first is before the installation of Acoustic Side Branch (ASB), and in the second case we consider the installation of said ASB in the standpipes of SRV. The analysis was performed with numerical experiments on a platform for computational fluid dynamics with virtual geometries previously designed based on the actual components of the reactor and steam system. The model to study is delimited by the top of the RPV, the steam dryer and a section of each of the four MSL with ten standpipes of SRV. With the obtained data and considering the mechanical-structural properties of the steam dryer material, we can evaluate the mechanical resistance to impacts by acoustic pressure load and its possible deformation or cracking. (Author)

  19. Stability analysis of a recycling circuit of a BWR type reactor. Theoretical study

    International Nuclear Information System (INIS)

    Salinas H, J.G.; Espinosa P, G.; Gonzalez M, V.M.

    2000-01-01

    The Technology, Regulation and Services Management of the National Commission of Nuclear Safety and Safeguards financed and in coordinate form with the I.P.H. Department of the Metropolitan Autonomous-Iztapalapa University developed the present project with the purpose of studying the effect of the recycling system on the linear stability of a BWR reactor whose reference central is the Laguna Verde power station. The present project forms part of a work series focused to the linear stability of the nuclear reactor of the Unit 1 at Laguna Verde power station. The components of the recycling system considered for the study of stability are the recycling external circuit (recycling pumps, valves) and the internal circuit (downcomer, jet pumps, lower full, driers, separators). The mathematical model is obtained applying mass balances and movement quantity in each one of the mentioned circuits. With respect to the nucleus model two regions are considered, the first one is made of a flow in one phase and the second one of a flow in two phases. For modelling the biphasic region it is considered homogenous flow. Generally it is studied the system behavior in the frequency domain starting from the transfer function applied to four operational states which correspond to the lower stability zone in the map power-flow of the Unit 1 of Laguna Verde power station. The Nyquist diagrams corresponding to each state as well as their characteristic frequency were determined. The results show that exists a very clear dependence of the power-flow relation on the stability of the system. It was found that the boiling length is an important parameter for the linear stability of the system. The obtained results show that the characteristic frequencies in unstability zones are similar to the reported data of the Unit 1 of the Laguna Verde power station in the event of power oscillations carried out in January 1995. (Author)

  20. 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

  1. Phased array concept for the ultrasonic inservice inspection of the spherical bottom of BWR-pressure vessels

    International Nuclear Information System (INIS)

    Brekow, G.; Wuestenberg, H.; Moehrle, W.; Schulz, E.

    1989-01-01

    The spherical bottom of BWR-pressure vessels contains holes for the nozzles of control rods and instrumentation. Up to now the detectable areas for the ultrasonic inspection are the accessible ligaments between the nozzles with an orientation parallel and transverse to the manipulator rails. Some licensing authorities demand an inspection technique capable of reliably detecting significant crack initiation in all critical areas near the cladding of the spherical inner surface. By order and in cooperation with the HEW we have developed a computer controlled equipment with two ultrasonic probes containing four linear arrays and a digitized A-scan storage for documentation and evaluation of inspection results. The manipulator guided probe movement in the paths between the nozzles of the spherical bottom is controlled by a computer program. This program determines for each array system and for each coupling position the beam angle as a function of the variable skewing angle to realize detection conditions suited to possible crack positions at the longitudinal, transverse and diagonal ligaments between the nozzles for control rods and instrumentation. (orig./HP)

  2. On-line fatigue monitoring system for reactor pressure vessel

    International Nuclear Information System (INIS)

    Tokunaga, K.; Sakai, A.; Aoki, T.; Ranganath, S.; Stevens, G.L.

    1994-01-01

    A workstation-based, on-line fatigue monitoring system for tracking fatigue usage applied to an operating boiling water reactor (BWR), Tsuruga Unit-1, is described. The system uses the influence function approach and determines component stresses using temperature, pressure, and flow rate data that are made available via signal taps from previously existing plant sensors. Using plant unique influence functions developed specifically for the feedwater nozzle location, the system calculates stresses as a function of time and computed fatigue usage. The analysis method used to compute fatigue usage complies with MITI Code Notification No.501. Fatigue usage results for an entire fuel cycle are presented and compared to assumed design basis events to confirm that actual plant thermal duty is significantly less severe than originally estimated in the design basis stress report. As a result, the system provides the technical basis to more accurately evaluate actual reactor conditions as well as the justification for plant life extension. (author)

  3. 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

  4. 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

  5. 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.)

  6. Simplified distributed parameters BWR dynamic model for transient and stability analysis

    International Nuclear Information System (INIS)

    Espinosa-Paredes, Gilberto; Nunez-Carrera, Alejandro; Vazquez-Rodriguez, Alejandro

    2006-01-01

    This paper describes a simplified model to perform transient and linear stability analysis for a typical boiling water reactor (BWR). The simplified transient model was based in lumped and distributed parameters approximations, which includes vessel dome and the downcomer, recirculation loops, neutron process, fuel pin temperature distribution, lower and upper plenums reactor core and pressure and level controls. The stability was determined by studying the linearized versions of the equations representing the BWR system in the frequency domain. Numerical examples are used to illustrate the wide application of the simplified BWR model. We concluded that this simplified model describes properly the dynamic of a BWR and can be used for safety analysis or as a first approach in the design of an advanced BWR

  7. Reactor power automatically controlling method and device for BWR type reactor

    International Nuclear Information System (INIS)

    Murata, Akira; Miyamoto, Yoshiyuki; Tanigawa, Naoshi.

    1997-01-01

    For an automatic control for a reactor power, when a deviation exceeds a predetermined value, the aimed value is kept at a predetermined value, and when the deviation is decreased to less than the predetermined value, the aimed value is increased from the predetermined value again. Alternatively, when a reactor power variation coefficient is decreased to less than a predetermine value, an aimed value is maintained at a predetermined value, and when the variation coefficient exceeds the predetermined value, the aimed value is increased. When the reactor power variation coefficient exceeds a first determined value, an aimed value is increased to a predetermined variation coefficient, and when the variation coefficient is decreased to less than the first determined value and also when the deviation between the aimed value and an actual reactor power exceeds a second determined value, the aimed value is maintained at a constant value. When the deviation is increased or when the reactor power variation coefficient is decreased, since the aimed value is maintained at predetermined value without increasing the aimed value, the deviation is not increased excessively thereby enabling to avoid excessive overshoot. (N.H.)

  8. BWR [boiling-water reactor] and PWR [pressurized-water reactor] off-normal event descriptions

    International Nuclear Information System (INIS)

    1987-11-01

    This document chronicles a total of 87 reactor event descriptions for use by operator licensing examiners in the construction of simulator scenarios. Events are organized into four categories: (1) boiling-water reactor abnormal events; (2) boiling-water reactor emergency events; (3) pressurized-water reactor abnormal events; and (4) pressurized-water reactor emergency events. Each event described includes a cover sheet and a progression of operator actions flow chart. The cover sheet contains the following general information: initial plant state, sequence initiator, important plant parameters, major plant systems affected, tolerance ranges, final plant state, and competencies tested. The progression of operator actions flow chart depicts, in a flow chart manner, the representative sequence(s) of expected immediate and subsequent candidate actions, including communications, that can be observed during the event. These descriptions are intended to provide examiners with a reliable, performance-based source of information from which to design simulator scenarios that will provide a valid test of the candidates' ability to safely and competently perform all licensed duties and responsibilities

  9. Design of a mixed recharge with MOX assemblies of greater relation of moderation for a BWR reactor; Diseno de una recarga mixta con ensambles MOX de mayor relacion de moderacion para un reactor BWR

    Energy Technology Data Exchange (ETDEWEB)

    Ramirez S, J.R.; Alonso V, G.; Palacios H, J. [ININ, Carretera Mexico-Toluca Km. 36.5, 52045 Estado de Mexico (Mexico)]. e-mail: jrrs@nuclear.inin.mx

    2004-07-01

    The study of the fuel of mixed oxides of uranium and plutonium (MOX) it has been topic of investigation in many countries of the world and those are even discussed in many places the benefits of reprocessing the spent fuel to extract the plutonium created during the irradiation of the fuel in the nuclear power reactors. At the moment those reactors that have been loaded partially with MOX fuel, are mainly of the type PWR where a mature technology has been achieved in some countries like they are France, Belgium and England, however the experience with reactors of the type BWR is more limited and it is continued studying the best way to introduce this type of fuel in BWRs, one of the main problems to introduce MOX in reactors BWR is the neutronic design of the same one, existing different concepts to introduce the plutonium in the assemblies of fuel and one of them is the one of increasing the relationship of moderation of the assemble. In this work a MOX fuel assemble design is presented and the obtained results so far in the ININ. These results indicate that the investigated concept has some exploitable advantages in the use of the MOX fuel. (Author)

  10. Reactor vessel using metal oxide ceramic membranes

    Science.gov (United States)

    Anderson, Marc A.; Zeltner, Walter A.

    1992-08-11

    A reaction vessel for use in photoelectrochemical reactions includes as its reactive surface a metal oxide porous ceramic membrane of a catalytic metal such as titanium. The reaction vessel includes a light source and a counter electrode. A provision for applying an electrical bias between the membrane and the counter electrode permits the Fermi levels of potential reaction to be favored so that certain reactions may be favored in the vessel. The electrical biasing is also useful for the cleaning of the catalytic membrane.

  11. 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

  12. 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

  13. 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

  14. 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)

  15. Feasibility of underwater welding of highly irradiated in-vessel components of boiling-water reactors: A literature review

    International Nuclear Information System (INIS)

    Lund, A.L.

    1997-11-01

    In February 1997, the U.S. Nuclear Regulatory Commission (NRC), Office of Nuclear Regulatory Research (RES), initiated a literature review to assess the state of underwater welding technology. In particular, the objective of this literature review was to evaluate the viability of underwater welding in-vessel components of boiling water reactor (BWR) in-vessel components, especially those components fabricated from stainless steels that are subjected to high neutron fluences. This assessment was requested because of the recent increased level of activity in the commercial nuclear industry to address generic issues concerning the reactor vessel and internals, especially those issues related to repair options. This literature review revealed a preponderance of general information about underwater welding technology, as a result of the active research in this field sponsored by the U.S. Navy and offshore oil and gas industry concerns. However, the literature search yielded only a limited amount of information about underwater welding of components in low-fluence areas of BWR in-vessel environments, and no information at all concerning underwater welding experiences in high-fluence environments. Research reported by the staff of the U.S. Department of Energy (DOE) Savannah River Site and researchers from the DOE fusion reactor program proved more fruitful. This research documented relevant experience concerning welding of stainless steel materials in air environments exposed to high neutron fluences. It also addressed problems with welding highly irradiated materials, and primarily attributed those problems to helium-induced cracking in the material. (Helium is produced from the neutron irradiation of boron, an impurity, and nickel.) The researchers found that the amount of helium-induced cracking could be controlled, or even eliminated, by reducing the heat input into the weld and applying a compressive stress perpendicular to the weld path

  16. 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)

  17. Fatigue evaluation in reactor vessel components

    International Nuclear Information System (INIS)

    Mattar Neto, Miguel; Miranda, Carlos A. de J.

    1994-01-01

    This paper presents a sequence of increasing complexity forms of evaluating fatigue damage of nuclear pressure vessel components caused by cycling loadings. Examples are included in order to illustrate such procedures. (author)

  18. BWR 90: The ABB advanced BWR design

    International Nuclear Information System (INIS)

    Haukeland, S.; Ivung, B.; Pedersen, T.

    1999-01-01

    ABB has two evolutionary advanced fight water reactors available today - the BWR 90 boiling water reactor and the System 80+ pressurised water reactor. The BWR 90 is based on the design, construction, commissioning and operation of the BWR 75 plants. The operation experience of the six plants of this advanced design has been very good. The average annual energy availability is above 90%, and the total power generation costs have been low. In the development of BWR 90 specific changes were introduced to the reference design, to adapt to technological progress, new safety requirements and to achieve cost savings. The thermal power rating of BWR 90 is 3800 MWth (providing a nominal 1374 MWe net), slightly higher dim that of the reference plant ABB Atom has taken advantage of margins gained using a new generation of its SVEA fuel to attain this power rating without major design modifications. The BWR 90 design was completed and offered to the TVO utility in Finland in 1991, as one of the contenders for the fifth Finnish nuclear power plant project. Thus, the design is available today for deployment in new plant projects. Utility views were incorporated through co-operation with the Finnish utility TVO, owner and operator of the two Olkiluoto plants of BWR 75 design. A review against the European Utility Requirement (EUR) set of requirements has been performed, since the design, in 1997, was selected by the EUR Steering Committee to be the first BWR to be evaluated against the EUR documents. The work is scheduled for completion in 1998. It will be the subject of an 'EUR Volume 3 Subset for BWR 90' document. ABB is continuing its BWR development work with the 'evolutionary' design BWR 90+. The primary design goal is to develop the BWR as a competitive option for the anticipated revival of the market for new nuclear plants beyond the turn of the century, as well as feeding ideas and inputs to the continuous modernisation efforts at operating plants. The development is

  19. Ex-Vessel corium coolability and steam explosion energetics in nordic light water reactors

    Energy Technology Data Exchange (ETDEWEB)

    Dinh, T.N.; Ma, W.M.; Karbojian, A.; Kudinov, P.; Tran, C.T.; Hansson, C.R. [Royal Institute of Technology (KTH), (Sweden)

    2008-03-15

    This report presents advances and insights from the KTH's study on corium pool heat transfer in the BWR lower head; debris bed formation; steam explosion energetics; thermal hydraulics and coolability in bottom-fed and heterogeneous debris beds. Specifically, for analysis of heat transfer in a BWR lower plenum an advanced threedimensional simulation tool was developed and validated, using a so-called effective convectivity approach and Fluent code platform. An assessment of corium retention and coolability in the reactor pressure vessel (RPV) lower plenum by means of water supplied through the Control Rod Guide Tube (CRGT) cooling system was performed. Simulant material melt experiments were performed in an intermediate temperature range (1300-1600K) on DEFOR test facility to study formation of debris beds in high and low subcooled water pools characteristic of in-vessel and ex-vessel conditions. Results of the DEFOR-E scoping experiments and related analyses strongly suggest that porous beds formed in ex-vessel from a fragmented high-temperature debris is far from homogeneous. Calculation results of bed thermal hydraulics and dryout heat flux with a two-dimensional thermal-hydraulic code give the first basis to evaluate the extent by which macro and micro inhomogeneity can enhance the bed coolability. The development and validation of a model for two-phase natural circulation through a heated porous medium and its application to the coolability analysis of bottom-fed beds enables quantification of the significant effect of dryout heat flux enhancement (by a factor of 80-160%) due to bottom coolant injection. For a qualitative and quantitative understanding of steam explosion, the SHARP system and its image processing methodology were used to characterize the dynamics of a hot liquid (melt) drop fragmentation and the volatile liquid (coolant) vaporization. The experimental results provide a basis to suggest that the melt drop preconditioning is instrumental to

  20. Ex-Vessel corium coolability and steam explosion energetics in nordic light water reactors

    International Nuclear Information System (INIS)

    Dinh, T.N.; Ma, W.M.; Karbojian, A.; Kudinov, P.; Tran, C.T.; Hansson, C.R.

    2008-03-01

    This report presents advances and insights from the KTH's study on corium pool heat transfer in the BWR lower head; debris bed formation; steam explosion energetics; thermal hydraulics and coolability in bottom-fed and heterogeneous debris beds. Specifically, for analysis of heat transfer in a BWR lower plenum an advanced threedimensional simulation tool was developed and validated, using a so-called effective convectivity approach and Fluent code platform. An assessment of corium retention and coolability in the reactor pressure vessel (RPV) lower plenum by means of water supplied through the Control Rod Guide Tube (CRGT) cooling system was performed. Simulant material melt experiments were performed in an intermediate temperature range (1300-1600K) on DEFOR test facility to study formation of debris beds in high and low subcooled water pools characteristic of in-vessel and ex-vessel conditions. Results of the DEFOR-E scoping experiments and related analyses strongly suggest that porous beds formed in ex-vessel from a fragmented high-temperature debris is far from homogeneous. Calculation results of bed thermal hydraulics and dryout heat flux with a two-dimensional thermal-hydraulic code give the first basis to evaluate the extent by which macro and micro inhomogeneity can enhance the bed coolability. The development and validation of a model for two-phase natural circulation through a heated porous medium and its application to the coolability analysis of bottom-fed beds enables quantification of the significant effect of dryout heat flux enhancement (by a factor of 80-160%) due to bottom coolant injection. For a qualitative and quantitative understanding of steam explosion, the SHARP system and its image processing methodology were used to characterize the dynamics of a hot liquid (melt) drop fragmentation and the volatile liquid (coolant) vaporization. The experimental results provide a basis to suggest that the melt drop preconditioning is instrumental to the

  1. Simulation of the flow obstruction of a jet pump in a BWR reactor with the code RELAP/SCDAPSIM; Simulacion de la obstruccion de flujo de una bomba jet en un reactor BWR con el codigo RELAP/SCDAPSIM

    Energy Technology Data Exchange (ETDEWEB)

    Cardenas V, J.; Filio L, C., E-mail: jaime.cardenas@cnsns.gob.mx [Comision Nacional de Seguridad Nuclear y Salvaguardias, Dr. Jose M. Barragan 779, Col. Narvarte, 03020 Ciudad de Mexico (Mexico)

    2016-09-15

    This work simulates the flow obstruction of a jet pump in one of the recirculation loops of a nuclear power plant with a reactor of type BWR at 100% of operating power, in order to analyze the behavior of the total flow of the refrigerant passing through the reactor core, the total flow in each recirculation loop of the reactor, together with the 10 jet pumps of each loop. The behavior of the power and the reactivity insertion due to the change of the refrigerant flow pattern is also analyzed. The simulation was carried out using the RELAP/SCDAPSIM version 3.5 code, using a reactor model with 10 jet pumps in each recirculation loop and a core consisting of 6 radial zones and 25 axial zones. The scenario postulates the flow obstruction in a jet pump in a recirculation loop A when the reactor operates at 100% rated power, causing a change in the total flow of refrigerant in the reactor core, leading to a decrease in power. Once the reactor conditions are established to its new power, the operator tries to recover the nominal power using the flow control valve of the recirculation loop A, opening stepwise as a strategy to safely recover the reactor power. In this analysis is assumed that the intention of the nuclear plant operator is to maintain the operation of the reactor during the established cycle. (Author)

  2. Assessment of two BWR accident management strategies

    International Nuclear Information System (INIS)

    Hodge, S.A.; Petek, M.

    1994-01-01

    Candidate mitigative strategies for the management of in-vessel events during the late phase (after-core degradation has occurred) of postulated boiling water reactor (BWR) severe accidents were considered at Oak Ridge National Laboratory (ORNL) during 1990. The identification of new strategies was subject to the constraint that they should, to the maximum extent possible, make use of the existing equipment and water resources of the BWR facilities, and not require major equipment modifications or additions. As a result of this effort, two of these candidate strategies were recommended for further assessment. The first was a strategy for containment flooding to maintain the core and structural debris within the reactor vessel in the event that vessel injection cannot be restored to terminate a severe accident sequence. The second strategy pertained to the opposite case, for which vessel injection would be restored after control blade melting had begun; its purpose was to provide an injection source of borated water at the concentration necessary to preclude criticality upon recovering a damaged BWR core. Assessments of these two strategies were performed during 1991 and this paper provides a discussion of the motivation for and purpose of these strategies, and the potential for their success. ((orig.))

  3. Flow model study of 'Monju' reactor vessel

    International Nuclear Information System (INIS)

    Miyaguchi, Kimihide

    1980-01-01

    In the case of designing the structures in nuclear reactors, various problems to be considered regarding thermo-hydrodynamics exist, such as the distribution of flow quantity and the pressure loss in reactors and the thermal shock to inlet and outlet nozzles. In order to grasp the flow characteristics of coolant in reactors, the 1/2 scale model of the reactor structure of ''Monju'' was attached to the water flow testing facility in the Oarai Engineering Center, and the simulation experiment has been carried out. The flow characteristics in reactors clarified by experiment and analysis so far are the distribution of flow quantity between high and low pressure regions in reactors, the distribution of flow quantity among flow zones in respective regions of high and low pressure, the pressure loss in respective parts in reactors, the flow pattern and the mixing effect of coolant in upper and lower plenums, the effect of the twisting angle of inlet nozzles on the flow characteristics in lower plenums, the effect of internal cylinders on the flow characteristics in upper plenums and so on. On the basis of these test results, the improvement of the design of structures in reactors was made, and the confirmation test on the improved structures was carried out. The testing method, the calculation method, the test results and the reflection to the design of actual machines are described. (Kako, I.)

  4. Design of a mixed recharge with MOX assemblies of greater relation of moderation for a BWR reactor

    International Nuclear Information System (INIS)

    Ramirez S, J.R.; Alonso V, G.; Palacios H, J.

    2004-01-01

    The study of the fuel of mixed oxides of uranium and plutonium (MOX) it has been topic of investigation in many countries of the world and those are even discussed in many places the benefits of reprocessing the spent fuel to extract the plutonium created during the irradiation of the fuel in the nuclear power reactors. At the moment those reactors that have been loaded partially with MOX fuel, are mainly of the type PWR where a mature technology has been achieved in some countries like they are France, Belgium and England, however the experience with reactors of the type BWR is more limited and it is continued studying the best way to introduce this type of fuel in BWRs, one of the main problems to introduce MOX in reactors BWR is the neutronic design of the same one, existing different concepts to introduce the plutonium in the assemblies of fuel and one of them is the one of increasing the relationship of moderation of the assemble. In this work a MOX fuel assemble design is presented and the obtained results so far in the ININ. These results indicate that the investigated concept has some exploitable advantages in the use of the MOX fuel. (Author)

  5. 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

  6. 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.

  7. 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

  8. 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)

  9. Design and development of weld inspection manipulator for reactor pressure vessel of TAPS-1

    International Nuclear Information System (INIS)

    Chatterjee, H.; Singh, J.P.; Ranjon, R.; Kulkarni, M.P.; Patel, R.J.

    2013-01-01

    The reactor pressure vessel (RPV) of TAPS-1 BWR contains six longitudinal and four circumferential welds. Periodical in-service inspection of these weld joints has been a regulatory issue pending for long. In the 22 nd refuelling outage in July 2012 the inspection of L1-1, L1-2 longitudinal welds as well as their junctions with C1 circumferential weld were proposed to be done using ultrasonic technique. Approaching these welds from OD side of the RPV is a difficult and tedious task. Therefore it was decided to examine these welds from ID side of the RPV by filling the cavity with water and approaching the RPV from top. No technology was locally available to take the probes at a depth of 10-12 m under water. NPCIL approached RTD, BARC to develop an underwater manipulator to accomplish this task. RTD took up this work as a challenge and came out with the design of manipulator. The weld inspection manipulator (WIM) was fabricated on a war foot basis, tested and successfully implemented in the reactor for the first time in TAPS history. The entire activity was completed in three months time. This article gives the details of design, manufacturing, performance testing, qualification trials and implementation of WIM in the reactor. Ultrasonic testing techniques were developed by QAD, BARC which are not covered in this article. (author)

  10. Analysis of the accident at Fukushima Daiichi nuclear power plant in an A BWR reactor

    International Nuclear Information System (INIS)

    Escorcia O, D.; Salazar S, E.

    2016-09-01

    The present work aims to recreate the accident occurred at the Fukushima Daiichi nuclear power plant in Japan on March 11, 2011, making use of an academic simulator of forced circulation of the A BWR reactor provided by the IAEA to know the scope of this simulator. The simulator was developed and distributed by the IAEA for academic purposes and contains the characteristics and general elements of this reactor to be able to simulate transients and failures of different types, allowing also to observe the general behavior of the reactor, as well as several phenomena and present systems in the same. Is an educational tool of great value, but it does not have a scope that allows the training of plant operators. To recreate the conditions of the Fukushima accident in the simulator, we first have to know what events led to this accident, as well as the actions taken by operators and managers to reduce the consequences of this accident; and the sequence of events that occurred during the course of the accident. Differences in the nuclear power plant behavior are observed and interpreted throughout the simulation, since the Fukushima plant technology and the simulator technology are not the same, although they have several elements in common. The Fukushima plant had an event that by far exceeded the design basis, which triggered in an accident that occurred in the first place by a total loss of power supply, followed by the loss of cooling systems, causing a level too high in temperature, melting the core and damaging the containment accordingly, allowing the escape of hydrogen and radioactive material. As a result of the simulation, was determined that the scope of the IAEA academic simulator reaches the entrance of the emergency equipment, so is able to simulate almost all the events occurred at the time of the earthquake and the arrival of the tsunami in the nuclear power plant of Fukushima Daiichi. However, due to its characteristics, is not able to simulate later

  11. BWR emergency procedure guidelines

    International Nuclear Information System (INIS)

    Post, J.S.; Karner, E.F.; Stratman, R.A.

    1984-01-01

    This chapter describes plans for dealing with reactor accidents developed by the Boiling Water Reactor (BWR) Owners' Group in response to post-Three Mile Island US NRC requirements. The devised Emergency Procedure Guidelines (EPGs), applicable to all BWRs, are symptom-based rather than event-based. According to the EPGs, the operator does not need to identify what event is occurring in the plant in order to decide what action to take, but need only observe the symptoms (values and trends of key control parameters) which exist and take appropriate action to control these symptoms. The original objective was to provide reactor operator guidance in responding to a small break loss-of-coolant accident (LOCA), but subsequent revisions have included other types of reactor accidents. Topics considered include the reactor pressure vessel (RPV) control guideline, the primary containment control guideline, the secondary containment control guideline, the radioactivity release control guideline, multiple failures vs. the design basis, safe limits vs. technical specifications, the technical status, licensing, and implementation. The EPGs are based upon maintaining both adequate core cooling and primary containment integrity

  12. 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

  13. Simulation of the flow obstruction of a jet pump in a BWR reactor with the code RELAP/SCDAPSIM

    International Nuclear Information System (INIS)

    Cardenas V, J.; Filio L, C.

    2016-09-01

    This work simulates the flow obstruction of a jet pump in one of the recirculation loops of a nuclear power plant with a reactor of type BWR at 100% of operating power, in order to analyze the behavior of the total flow of the refrigerant passing through the reactor core, the total flow in each recirculation loop of the reactor, together with the 10 jet pumps of each loop. The behavior of the power and the reactivity insertion due to the change of the refrigerant flow pattern is also analyzed. The simulation was carried out using the RELAP/SCDAPSIM version 3.5 code, using a reactor model with 10 jet pumps in each recirculation loop and a core consisting of 6 radial zones and 25 axial zones. The scenario postulates the flow obstruction in a jet pump in a recirculation loop A when the reactor operates at 100% rated power, causing a change in the total flow of refrigerant in the reactor core, leading to a decrease in power. Once the reactor conditions are established to its new power, the operator tries to recover the nominal power using the flow control valve of the recirculation loop A, opening stepwise as a strategy to safely recover the reactor power. In this analysis is assumed that the intention of the nuclear plant operator is to maintain the operation of the reactor during the established cycle. (Author)

  14. Innovative inspection system for reactor pressure vessels

    International Nuclear Information System (INIS)

    Mertens, K.; Trautmann, H.

    1999-01-01

    The versatile, compact and modern underwater systems described, the DELPHIN manipulators and MIDAS submarines, are innovative systems enabling RPV inspections at considerably reduced efforts and time, thus reducing the total time required for ISI of reactors. (orig./CB) [de

  15. Evaluation of the radial design of fuel cells in an operation cycle of a BWR reactor; Evaluacion del diseno radial de celdas de combustible en un ciclo de operacion de un reactor BWR

    Energy Technology Data Exchange (ETDEWEB)

    Gonzalez C, J.; Martin del Campo M, C. [Laboratorio de Analisis en Ingenieria de Reactores Nucleares, Facultad de Ingenieria, UNAM, Paseo Cuauhnahuac 8532, Jiutepec, Morelos (Mexico)]. e-mail: jgco@ver.megared.net.mx

    2003-07-01

    This work is continuation of one previous in the one that the application of the optimization technique called Tabu search to the radial design of fuel cells of boiling water reactors (BWR, Boiling Water Reactor) is presented. The objective function used in the optimization process only include neutron parameters (k-infinite and peak of radial power) considering the cell at infinite media. It was obtained to reduce the cell average enrichment completing the characteristics of reactivity of an original cell. The objective of the present work is to validate the objective function that was used for the radial design of the fuel cell (test cell), analyzing the operation of a one cycle of the reactor in which fuels have been fresh recharged that contain an axial area with the nuclear database of the cell designed instead of the original cell. For it is simulated it with Cm-Presto the cycle 10 of the reactor operation of the Unit 1 of the Nuclear Power station of Laguna Verde (U1-CNLV). For the cycle evaluation its were applied so much the simulation with the Haling strategy, as the simulation of the one cycle with control rod patterns and they were evaluated the energy generation and several power limits and reactivity that are used as design parameters in fuel reloads of BWR reactors. The results at level of an operation cycle of the reactor, show that the objective function used in the optimization and radial design of the cell is adequate and that it can induce to one good use of the fuel. (Author)

  16. Evaluation of the cracking by stress corrosion in nuclear reactor environments type BWR

    International Nuclear Information System (INIS)

    Arganis J, C. R.

    2010-01-01

    The stress corrosion cracking susceptibility was studied in sensitized, solution annealed 304 steel, and in 304-L welded with a heat treatment that simulated the radiation induced segregation, by the slow strain rate test technique, in a similar environment of a boiling water reactor (BWR), 288 C, 8 MPa, low conductivity and a electrochemical corrosion potential near 200 mV. vs. standard hydrogen electrode (She). The electrochemical noise technique was used for the detection of the initiation and propagation of the cracking. The steels were characterized by metallographic studies with optical and scanning electronic microscopy and by the electrochemical potentiodynamic reactivation of single loop and double loop. In all the cases, the steels present delta ferrite. The slow strain rate tests showed that the 304 steel in the solution annealed condition is susceptible to transgranular stress corrosion cracking (TGSCC), such as in a normalized condition showed granulated. In the sensitized condition the steel showed intergranular stress corrosion cracking, followed by a transition to TGSCC. The electrochemical noise time series showed that is possible associated different time sequences to different modes of cracking and that is possible detect sequentially cracking events, it is means, one after other, supported by the fractographic studies by scanning electron microscopy. The parameter that can distinguish between the different modes of cracking is the re passivation rate, obtained by the current decay rate -n- in the current transients. This is due that the re passivation rate is a function of the microstructure and the sensitization. Other statistic parameters like the localized index, Kurtosis, Skew, produce results that are related with mixed corrosion. (Author)

  17. 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

  18. 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

  19. Calculation of releases of radioactive materials in gaseous and liquid effluents from boiling water reactors (BWR-GALE Code)

    International Nuclear Information System (INIS)

    Bangart, R.L.; Bell, L.G.; Boegli, J.S.; Burke, W.C.; Lee, J.Y.; Minns, J.L.; Stoddart, P.G.; Weller, R.A.; Collins, J.T.

    1978-12-01

    The calculational procedures described in the report reflect current NRC staff practice. The methods described will be used in the evaluation of applications for construction permits and operating licenses docketed after January 1, 1979, until this NUREG is revised as a result of additional staff review. The BWR-GALE (Boiling Water Reactor Gaseous and Liquid Effluents) Code is a computerized mathematical model for calculating the release of radioactive material in gaseous and liquid effluents from boiling water reactors (BWRs). The calculations are based on data generated from operating reactors, field tests, laboratory tests, and plant-specific design considerations incorporated to reduce the quantity of radioactive materials that may be released to the environment

  20. 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

  1. 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

  2. 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

  3. BWR condensate filtration studies

    International Nuclear Information System (INIS)

    Wilson, J.A.; Pasricha, A.; Rekart, T.E.

    1993-09-01

    Poor removal of particulate corrosion products (especially iron) from condensate is one of the major problems in BWR systems. The presence of activated corrosion products creates ''hot spots'' and increases piping dose rates. Also, fuel efficiency is reduced and the risk of fuel failure is increased by the deposit of corrosion products on the fuel. Because of these concerns, current EPRI guidelines call for a maximum of 2 ppb of iron in the reactor feedwater with a level of 0.5 ppb being especially desirable. It has become clear that conventional deep bed resins are incapable of meeting these levels. While installation of prefilter systems is an option, it would be more economical for plants with naked deep beds to find an improved bead resin for use in existing systems. BWR condensate filtration technologies are being tested on a condensate side stream at Hope Creek Nuclear Generating Station. After two years of testing, hollow fiber filters (HFF) and fiber matrix filters (FMF), and low crosslink cation resin, all provide acceptable results. The results are presented for pressure drop, filtration efficiency, and water quality measurements. The costs are compared for backwashable non-precoat HFF and FMF. Results are also presented for full deep bed vessel tests of the low crosslink cation resin

  4. Fuel design with low peak of local power for BWR reactors with increased nominal power

    International Nuclear Information System (INIS)

    Perusquia C, R.; Montes, J.L.; Hernandez, J.L.; Ortiz, J.J.; Castillo, A.

    2006-01-01

    The Federal Commission of Electricity recently announcement the beginning of the works related with the increase of the power to 120% of the original nominal one in the Boiling Water Reactors (BWR) of the Laguna Verde Central (CLV): In the National Institute of Nuclear Research (ININ) are carried out studies of the impact on the design of the recharge of derived fuel of this increase. One of the main effects of the power increase type that it is promoting, is the increment of the flow of generated vapor, what takes, to a bigger fraction of vacuum in the core presenting increased values of the maximum fraction to the limit, so much of the ratio of lineal heat generation (XFLPD) as of the ratio of critic power (MFLCPR). In the made studies, it is found that these fractions rise lineally with the increase of the nominal power. Considering that the reactors of the CLV at the moment operate to 105% of the original nominal power, it would imply an increment of the order of 13.35% in the XFLPD and in the MFLCPR operating to a nominal power of 120% of the original one. This would propitiate bigger problems to design appropriately the fuel cycle and the necessity, almost unavoidable, of to resort to a fuel assembly type more advanced for the recharges of the cores. As option, in the ININ the feasibility of continuing using the same type of it fuel assembles that one has come using recently in the CLV, the type GE12 is analyzed. To achieve it was outlined to diminish the peak factor of local power (LPPF) of the power cells that compose the fuel recharge in 13.35%. It was started of a fuel design previously used in the recharge of the unit 1 cycle 12 and it was re-design to use it in the recharge design of the cycle 13 of the unit 1, considering an increase to 120% of the original power and the same requirements of cycle extension. For the re-design of the fuel assembly cell it was used the PreDiCeldas computer program developed in the ININ. It was able to diminish the LPPF

  5. 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

  6. 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.

  7. CFD Simulation of rigid venting of the containment of a BWR-5 Mark-II reactor; Simulacion CFD de los venteos rigidos de la contencion de un reactor BWR-5 Mark-II

    Energy Technology Data Exchange (ETDEWEB)

    Galindo G, I. F.; Vazquez B, A. K.; Velazquez E, L. [Instituto Nacional de Electricidad y Energias Limpias, Reforma 113, Col. Palmira, 62490 Cuernavaca, Morelos (Mexico); Tijerina S, F.; Tapia M, R., E-mail: francisco.tijerina@cfe.gob.mx [CFE, Central Nucleoelectrica Laguna Verde, Carretera Federal Cardel-Nautla Km 42.5, 91476 Municipio Alto Lucero, Veracruz (Mexico)

    2016-09-15

    In conditions of prolonged loss of external energy or a severe accident, venting to the atmosphere is an alternative to prevent overpressure and release of fission products from the primary containment of a nuclear reactor. Due to the importance of flow determination through rigid vents, a computational fluid dynamics (CFD) model is proposed to verify the capacity of rigid vents in the primary containment of a boiling water reactor (BWR) under different operating conditions (pressure, temperature and compositions of the fluids). The model predicts and provides detailed information on variables such as mass flow and velocity of the venting gases. In the proposed model the primary containment gas is vented to the atmosphere via rigid vents (pipes) from the dry and wet pit. Is assumed that the container is pressurized because is in a defined scenario, and at one point the venting is open and the gas released into the atmosphere. The objective is to characterize the flow and validate the CFD model for the overpressure conditions that occur in an accident such as a LOCA, Sbo, etc. The model is implemented with Ansys-Fluent general-purpose CFD software based on the geometry of the venting ducts of the containment of a BWR. The model is developed three-dimensional and resolves at steady state for compressible flow and includes the effects of the turbulence represented by the Reynolds stress model. The CFD results are compared with the values of a one-dimensional and isentropic model for compressible flow. The relative similarity of results leads to the conclusion that the proposed CFD model can help to predict the rigid venting capacity of the containment of a BWR, however more information is required for full validation of the proposed model. (Author)

  8. 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

  9. 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)

  10. 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

  11. Treatment of core components from nuclear power plants with PWR and BWR reactors - 16043

    International Nuclear Information System (INIS)

    Viermann, Joerg; Friske, Andreas; Radzuweit, Joerg

    2009-01-01

    During operation of a Nuclear Power Plant components inside the RPV get irradiated. Irradiation has an effect on physical properties of these components. Some components have to be replaced after certain neutron doses or respectively after a certain operating time of the plant. Such components are for instance water channels and control rods from Boiling Water Reactors (BWR) or control elements, poisoning elements and flow restrictors from Pressurized Water Reactors (PWR). Most of these components are stored in the fuel pool for a certain time after replacement. Then they have to be packaged for further treatment or for disposal. More than 25 years ago GNS developed a system for disposal of irradiated core components which was based on a waste container suitable for transport, storage and disposal of Intermediate Level Waste (ILW), the so-called MOSAIK R cask. The MOSAIK R family of casks is subject of a separate presentation at the ICEM 09 conference. Besides the MOSAIK R cask the treatment system developed by GNS comprised underwater shears to cut the components to size as well as different types of equipment to handle the components, the shears and the MOSAIK R casks in the fuel pool. Over a decade of experience it showed that this system although effective needed improvement for BWR plants where many water channels and control rods had to be replaced after a certain operating time. Because of the large numbers of components the time period needed to cut the components in the pool had a too big influence on other operational work like rearranging of fuel assemblies in the pool. The system was therefore further developed and again a suitable cask was the heart of the solution. GNS developed the type MOSAIK R 80 T, a cask that is capable to ship the unsegmented components with a length of approx. 4.5 m from the Power plants to an external treatment centre. This treatment centre consisting of a hot cell installation with a scrap shear, super-compactor and a heavy

  12. 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)

  13. 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

  14. 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

  15. 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)

  16. 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

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

    International Nuclear Information System (INIS)

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

    1983-01-01

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

  18. 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

  19. Summary of the OECD/NRC Boiling Water Reactor Turbine Trip Benchmark - Fifth Workshop (BWR-TT5)

    International Nuclear Information System (INIS)

    2003-01-01

    The reference problem chosen for simulation in a BWR is a Turbine Trip transient, which begins with a sudden Turbine Stop Valve (TSV) closure. The pressure oscillation generated in the main steam piping propagates with relatively little attenuation into the reactor core. The induced core pressure oscillation results in dramatic changes of the core void distribution and fluid flow. The magnitude of the neutron flux transient taking place in the BWR core is strongly affected by the initial rate of pressure rise caused by pressure oscillation and has a strong spatial variation. The correct simulation of the power response to the pressure pulse and subsequent void collapse requires a 3-D core modeling supplemented by 1-D simulation of the remainder of the reactor coolant system. A BWR TT benchmark exercise, based on a well-defined problem with complete set of input specifications and reference experimental data, has been proposed for qualification of the coupled 3-D neutron kinetics/thermal-hydraulic system transient codes. Since this kind of transient is a dynamically complex event with reactor variables changing very rapidly, it constitutes a good benchmark problem to test the coupled codes on both levels: neutronics/thermal-hydraulic coupling and core/plant system coupling. Subsequently, the objectives of the proposed benchmark are: comprehensive feedback testing and examination of the capability of coupled codes to analyze complex transients with coupled core/plant interactions by comparison with actual experimental data. The benchmark consists of three separate exercises: Exercise 1 - Power vs. Time Plant System Simulation with Fixed Axial Power Profile Table (Obtained from Experimental Data). Exercise 2 - Coupled 3-D Kinetics/Core Thermal-Hydraulic BC Model and/or 1-D Kinetics Plant System Simulation. Exercise 3 - Best-Estimate Coupled 3-D Core/Thermal-Hydraulic System Modeling. The purpose of this fifth workshop was to discuss the results from Phase III (best

  20. 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

  1. PWR reactor pressure vessel failure probabilities

    International Nuclear Information System (INIS)

    Dufresne, J.; Lanore, J.M.; Lucia, A.C.; Elbaz, J.; Brunnhuber, R.

    1980-05-01

    To evaluate the rupture probability of a LWR vessel a probabilistic method using the fracture mechanics under probabilistic form has been proposed previously, but it appears that more accurate evaluation is possible. In consequence a joint collaboration agreement signed in 1976 between CEA, EURATOM, JRC Ispra and FRAMATOME set up and started a research program covering three parts: a computer code development, data acquisition and processing, and a support experimental program which aims at clarifying the most important parameters used in the COVASTOL computer code

  2. Determination of the neutron fluence in the welding of the 'Core shroud' of the BWR reactor core; Determinacion de la fluencia neutronica en las soldaduras del 'core shroud' del nucleo de un reactor BWR

    Energy Technology Data Exchange (ETDEWEB)

    Lucatero, M A; Xolocostli M, J V; Gomez T, A M; Palacios H, J C [ININ, 52750 Ocoyoacac, Estado de mexico (Mexico)

    2006-07-01

    With the purpose of defining the inspection frequency, in function of the embrittlement of the materials that compose the welding of the 'Core Shroud' or encircling of the core of a BWR type reactor, is necessary to know the neutron fluence received for this welding. In the work the calculated values of neutron fluence accumulated maxim (E > 1 MeV) during the first 8 operation cycles of the reactor are presented. The calculations were carried out according to the NRC Regulatory Guide 1.190, making use of the DORT code, which solves the transport equation in discreet ordinate in two dimensions (xy, r{theta}, and rz). The results in 3D were obtained applying the Synthesis method according to the guide before mentioned. Results are presented for the horizontal welding H3, H4, and H5, showing the corresponding curves to the fluence accumulated to the cycle 8 and a projection for the cycle 14 is presented. (Author)

  3. Influence of gamma irradiation on the deterioration of reactor pressure vessel materials and on reactor dosimetry measurements. Final report

    International Nuclear Information System (INIS)

    Boehmer, B.; Konheiser, J.; Kumpf, H.; Noack, K.; Vladimirov, P.

    2002-10-01

    Radiation embrittlement of pressure vessel steel in mixed neutron-gamma fields is mostly determined by neutrons, but in some cases also by gamma-radiation. Depending on the reactor type, gamma radiation can influence evaluations of lead factors of surveillance specimens, effect the interpretation of results of irradiation experiments and finally, it can result in changed pressure vessel lifetime evaluations. The project aimed at the evaluation of the importance of gamma radiation for RPV steel damage for several types of light-water reactors. Absolute neutron and gamma fluence rate spectra had been calculated for the Russian PWR types VVER-440 and two core loading variants of VVER-1000, for a German 1300 MW PWR and a German 900 MW BWR. Based on the calculated spectra several flux integrals and radiation damage parameters were derived for the region of the azimuthal flux maxima in the mid-planes for different radial positions between core and biological shield, especially, at the inner and outer surfaces of the PV walls, at the (1/4)-PV-thickness and at the surveillance positions. Fissionable materials are often used as activation detectors for neutron fluence measurements. To get the real value the analysis demands to take into account the gamma induced fissions. Therefore, the part of these fissions in the total number of fissions was estimated for the detector reactions 237 Np(n,f) and 238 U(n,f) in the calculated neutron/gamma fields. It has been found that considerable corrections of the neutron fluence measurements can be necessary, especially in case of 238 U(n,f). Most of the calculations were performed using a three-dimensional synthesis of 2D/1D-flux distributions obtained by the S N -code DORT with the BUGLE-96T group cross-section library. (orig.) [de

  4. 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

  5. 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.)

  6. 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

  7. Thermal performance of an insulating structure for a reactor vessel

    International Nuclear Information System (INIS)

    Aranovitch, E.; Crutzen, S.; LeDet, M.; Denis, R.

    This report describes the installations used to test the HTGR reactor vessel insulating structure called ''Casali'' and details the experimental results in 3 groups: general experiments, systematic study, and technological experiments. The results obtained make it possible to satisfactorily predict the behavior of the structure in a practical application

  8. 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

  9. Assessment and management of ageing of major nuclear power plant components important to safety: BWR pressure vessel internals

    International Nuclear Information System (INIS)

    2005-10-01

    . The guidance reports are directed at technical experts from NPPs and from regulatory, plant design, manufacturing and technical support organizations dealing with specific plant components addressed in the reports. The report addresses the reactor pressure vessel internals in BWRs. Maintaining the structural integrity of these reactor pressure vessel internals throughout NPP service life, in spite of several ageing mechanisms, is essential for plant safety

  10. 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)

  11. 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

  12. 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.)

  13. 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.)

  14. 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)

  15. 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)

  16. 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.

  17. Integral reactor vessel related to power reactor safety

    International Nuclear Information System (INIS)

    Widart, J.; Scailteur, A.

    1978-01-01

    Integral design applied to PWR pressure vessels allows to reach a high level of safety because: 1) it presents a better balance of the material in the geometry, resulting in an improved stress level (mainly faulted condition loadings); 2) location and geometry of the welds are designed in order to get a very sound pressure boundary of the upper part of the vessel; 3) the new location and geometry of the welds allow an easy ISI in such a way that ambiguity surrounding defect size or locaton is practically suppressed. (author)

  18. Genusa Bepu methodologies for the safety analysis of BWRs; Metodologias Bepu de Genusa para el analisis de seguridad de reactores BWR

    Energy Technology Data Exchange (ETDEWEB)

    Trueba, M.; Garcia, J.; Goodson, C.; Ibarra, L.

    2016-08-01

    This article describes the BEPU methodologies developed by General Electric-Hitachi (GEH) for the evaluation of the BWR reactor safety analysis based on the TRACG best-estimate code. These methodologies are applicable to a wide range of events, operational transients (AOO), anticipated transients without scram (ATWS), loss of coolant accidents (LOCA) and instability events; to different BWR types operating commercially. General Electric (GE( designs and other vendors, including Generation III+ESBWR; to the new operation strategies, and to all types of BWR fuel. Their application achieves, among other benefits, a better understanding of the overall plant response and an improvement in margins to the operating limits; thus, the increase of flexibility in reactor operation and reduction in generation costs. (Author)

  19. 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

  20. Advanced boiling water reactor

    International Nuclear Information System (INIS)

    Nishimura, N.; Nakai, H.; Ross, M.A.

    1999-01-01

    In the Boiling Water Reactor (BWR) system, steam generated within the nuclear boiler is sent directly to the main turbine. This direct cycle steam delivery system enables the BWR to have a compact power generation building design. Another feature of the BWR is the inherent safety that results from the negative reactivity coefficient of the steam void in the core. Based on the significant construction and operation experience accumulated on the BWR throughout the world, the ABWR was developed to further improve the BWR characteristics and to achieve higher performance goals. The ABWR adopted 'First of a Kind' type technologies to achieve the desired performance improvements. The Reactor Internal Pump (RIP), Fine Motion Control Rod Drive (FMCRD), Reinforced Concrete Containment Vessel (RCCV), three full divisions of Emergency Core Cooling System (ECCS), integrated digital Instrumentation and Control (I and C), and a high thermal efficiency main steam turbine system were developed and introduced into the ABWR. (author)

  1. Use of superheated steam to anneal the reactor pressure vessel

    International Nuclear Information System (INIS)

    Porowski, J.S.

    1994-01-01

    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 degrees 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 degrees 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

  2. 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)

  3. Updating of the costs of the nuclear fuels of the equilibrium reloading of the A BWR and EPR reactors

    International Nuclear Information System (INIS)

    Ortega C, R.F.

    2008-01-01

    In the last two and a half years, the price of the uranium in the market spot has ascended of US$20.00 dollars by lb U 3O 8 in January, 2005 to a maximum of US$137.00 dollars by Ib U 3 O 8 by the middle of 2007. At the moment this price has been stabilized in US$90.00 dollars by Ib U 3 O 8 such for the market spot, like for the long term contracts. In this work the reasons of this increment are analyzed, as well as their impact in the fuel prices of the balance recharge of the advanced reactors of boiling water (A BWR) and of the advanced water at pressure reactors (EPR). (Author)

  4. Determination of the neutron fluence in the welding of the 'Core shroud' of the BWR reactor core

    International Nuclear Information System (INIS)

    Lucatero, M.A.; Xolocostli M, J.V.; Gomez T, A.M.; Palacios H, J.C.

    2006-01-01

    With the purpose of defining the inspection frequency, in function of the embrittlement of the materials that compose the welding of the 'Core Shroud' or encircling of the core of a BWR type reactor, is necessary to know the neutron fluence received for this welding. In the work the calculated values of neutron fluence accumulated maxim (E > 1 MeV) during the first 8 operation cycles of the reactor are presented. The calculations were carried out according to the NRC Regulatory Guide 1.190, making use of the DORT code, which solves the transport equation in discreet ordinate in two dimensions (xy, rΘ, and rz). The results in 3D were obtained applying the Synthesis method according to the guide before mentioned. Results are presented for the horizontal welding H3, H4, and H5, showing the corresponding curves to the fluence accumulated to the cycle 8 and a projection for the cycle 14 is presented. (Author)

  5. 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

  6. 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

  7. Simulation in 3 dimensions of a cycle 18 months for an BWR type reactor using the Nod3D program; Simulacion en 3 dimensiones de un ciclo de 18 meses para un reactor BWR usando el programa Nod3D

    Energy Technology Data Exchange (ETDEWEB)

    Hernandez, N.; Alonso, G. [ININ, A.P. 18-1027, 11801 Mexico D.F. (Mexico)]. E-mail: nhm@nuclear.inin.mx; Valle, E. del [IPN, ESFM, 07738 Mexico D.F. (Mexico)

    2004-07-01

    The development of own codes that you/they allow the simulation in 3 dimensions of the nucleus of a reactor and be of easy maintenance, without the consequent payment of expensive use licenses, it can be a factor that propitiates the technological independence. In the Department of Nuclear Engineering (DIN) of the Superior School of Physics and Mathematics (ESFM) of the National Polytechnic Institute (IPN) a denominated program Nod3D has been developed with the one that one can simulate the operation of a reactor BWR in 3 dimensions calculating the effective multiplication factor (kJJ3, as well as the distribution of the flow neutronic and of the axial and radial profiles of the power, inside a means of well-known characteristics solving the equations of diffusion of neutrons numerically in stationary state and geometry XYZ using the mathematical nodal method RTN0 (Raviart-Thomas-Nedelec of index zero). One of the limitations of the program Nod3D is that it doesn't allow to consider the burnt of the fuel in an independent way considering feedback, this makes it in an implicit way considering the effective sections in each step of burnt and these sections are obtained of the code Core Master LEND. However even given this limitation, the results obtained in the simulation of a cycle of typical operation of a reactor of the type BWR are similar to those reported by the code Core Master LENDS. The results of the keJ - that were obtained with the program Nod3D they were compared with the results of the code Core Master LEND, presenting a difference smaller than 0.2% (200 pcm), and in the case of the axial profile of power, the maxim differs it was of 2.5%. (Author)

  8. 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)

  9. Peach Bottom transient analysis with BWR TRACB02

    International Nuclear Information System (INIS)

    Alamgir, M.; Sutherland, W.A.

    1984-01-01

    TRAC calculations have been performed for a Turbine Trip transient (TT1) in the Peach Bottom BWR power plant. This study is a part of the qualification of the BWR-TRAC code. The simulation is aimed at reproducing the observed thermal hydraulic behavior in a pressurization transient. Measured core power is an input to the calculation. Comparison with data show the code reasonably well predicts the generation and propagation of the pressure waves in the main steam line and associated pressurization of the reactor vessel following the closure of the turbine stop valve

  10. A practical methodology of radiological protection for the reduction of hot particles in BWR type reactors; Una metodologia practica de proteccion radiologica para la reduccion de particulas calientes en reactores BWR

    Energy Technology Data Exchange (ETDEWEB)

    Alvarez G, G [Comision Federal de Electricidad, Gerencia del Proyecto Nucleoelectrico Laguna Verde, Disciplina de Fisica Aplicada (Mexico)

    1991-07-01

    The purpose of this work, in general form, is to describe a practical method for reduction of hot particles generated as consequence of the operational activities of BWR nuclear reactors. This methodology provides a description of the localizations and/or probable activities of finding particles highly radioactive denominated hot particles. For this purpose it was developed a strategy based on the decontamination lineaments, as well as the manipulation, gathering, registration, contention, documentation, control and final disposition of the hot particles. In addition, some recommendations are reiterated and alternative, in order to gathering the hot particles in a dynamic way given to the activities of the personal occupationally exposed in highly radioactive areas. The structure of the methodology of hot particles is supported in the radiological controls based on the Code of Federal Regulation 10 CFR 20 as well as the applicable regulatory documents. It provides an idea based on administrative controls of radiological protection, in order to suggesting the responsibilities and necessary directing for the control of the hot particles required in nuclear plants of the BWR type. (author)

  11. Application of the MOVE algorithm for the identification of reduced order models of a core of a BWR type reactor; Aplicacion del algoritmo MOVE para la identificacion de modelos de orden reducido del nucleo de un reactor tipo BWR

    Energy Technology Data Exchange (ETDEWEB)

    Victoria R, M.A.; Morales S, J.B. [UNAM, DEPFI, Campus Morelos, en IMTA Jiutepec, Morelos (Mexico)]. e-mail: angelvr@gmail.com

    2005-07-01

    Presently work is applied the modified algorithm of the ellipsoid of optimal volume (MOVE) to a reduced order model of 5 differential equations of the core of a boiling water reactor (BWR) with the purpose of estimating the parameters that model the dynamics. The viability is analyzed of carrying out an analysis that calculates the global dynamic parameters that determine the stability of the system and the uncertainty of the estimate. The modified algorithm of the ellipsoid of optimal volume (MOVE), is a method applied to the parametric identification of systems, in particular to the estimate of groups of parameters (PSE for their initials in English). It is looked for to obtain the ellipsoid of smaller volume that guarantees to contain the real value of the parameters of the model. The PSE MOVE is a recursive identification method that can manage the sign of noise and to ponder it, the ellipsoid represents an advantage due to its easy mathematical handling in the computer, the results that surrender are very useful for the design of Robust Control since to smaller volume of the ellipsoid, better is in general the performance of the system to control. The comparison with other methods presented in the literature to estimate the reason of decline (DR) of a BWR is presented. (Author)

  12. 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

  13. 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

  14. Application of reliability techniques to prioritize BWR [boiling water reactor] recirculation loop welds for in-service inspection

    International Nuclear Information System (INIS)

    Holman, G.S.

    1989-12-01

    In January 1988 the US Nuclear Regulatory Commission issued Generic Letter 88-01 together with NUREG-0313, Revision 2, ''Technical Report on Material Selection and Processing Guidelines for BWR Coolant Pressure Boundary Piping,'' to implement NRC long-range plans for addressing the problem of stress corrosion cracking in boiling water reactor piping. NUREG-0313 presents guidelines for categorizing BWR pipe welds according to their SCC condition (e.g., presence of known cracks, implementation of measures for mitigating SCC) as well as recommended inspection schedules (e.g., percentage of welds inspected, inspection frequency) for each weld category. NUREG-0313 does not, however, specify individual welds to be inspected. To address this issue, the Lawrence Livermore National Laboratory developed two recommended inspection samples for welds in a typical BWR recirculation loop. Using a probabilistic fracture mechanics model, LLNL prioritized loop welds on the basis of estimated leak probabilities. The results of this evaluation indicate that riser welds and bypass welds should be given priority attention over other welds. Larger-diameter welds as a group can be considered of secondary importance compared to riser and bypass welds. A ''blind'' comparison between the probability-based inspection samples and data from actual field inspections indicated that the probabilistic analysis generally captured the welds which the field inspections identified as warranting repair or replacement. Discrepancies between the field data and the analytic results can likely be attributed to simplifying assumptions made in the analysis. The overall agreement between analysis and field experience suggests that reliability techniques -- when combined with historical experience -- represent a sound technical basis on which to define meaningful weld inspection programs. 13 refs., 8 figs., 5 tabs

  15. 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)

  16. 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

  17. 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

  18. 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)

  19. Phased array concept for the ultrasonic inservice inspection at the spherical bottom of BWR-pressure vessels

    International Nuclear Information System (INIS)

    Brekow, G.; Wuestenberg, H.; Moehrle, W.; Schulz, E.

    1987-01-01

    The required enhancement of the integrity assessment of perforated reactor vessel base plates has been achieved by a phased-array concept. This technique improvement is based on the fact that the quantity of individual resonant components is reduced whilst increasing the amount of web regions which fall into the sonic range of the pivoted detector due to the larger apex angle scope which the phased array concept provides. A mathematical model concept was initially developed to determine the acoustic irradiation angle and squiat angle ranges to be detected by the phased-array scanner. A prototype of this device has been constructed and tested with a steel sample possessing different perforations and experimental reflectors in order to assess and optimize the new system. The results of these investigations are presented together with those of an application at the nuclear power station in Brunsbuettel. (orig./DG) [de

  20. 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

  1. Simulation in 3 dimensions of a cycle 18 months for an BWR type reactor using the Nod3D program

    International Nuclear Information System (INIS)

    Hernandez, N.; Alonso, G.; Valle, E. del

    2004-01-01

    The development of own codes that you/they allow the simulation in 3 dimensions of the nucleus of a reactor and be of easy maintenance, without the consequent payment of expensive use licenses, it can be a factor that propitiates the technological independence. In the Department of Nuclear Engineering (DIN) of the Superior School of Physics and Mathematics (ESFM) of the National Polytechnic Institute (IPN) a denominated program Nod3D has been developed with the one that one can simulate the operation of a reactor BWR in 3 dimensions calculating the effective multiplication factor (kJJ3, as well as the distribution of the flow neutronic and of the axial and radial profiles of the power, inside a means of well-known characteristics solving the equations of diffusion of neutrons numerically in stationary state and geometry XYZ using the mathematical nodal method RTN0 (Raviart-Thomas-Nedelec of index zero). One of the limitations of the program Nod3D is that it doesn't allow to consider the burnt of the fuel in an independent way considering feedback, this makes it in an implicit way considering the effective sections in each step of burnt and these sections are obtained of the code Core Master LEND. However even given this limitation, the results obtained in the simulation of a cycle of typical operation of a reactor of the type BWR are similar to those reported by the code Core Master LENDS. The results of the keJ - that were obtained with the program Nod3D they were compared with the results of the code Core Master LEND, presenting a difference smaller than 0.2% (200 pcm), and in the case of the axial profile of power, the maxim differs it was of 2.5%. (Author)

  2. Experimental and analytical studies for a BWR nuclear reactor building. Evaluation of soil-structure interaction behaviour

    International Nuclear Information System (INIS)

    Mizuno, N.; Tsushima, Y.

    1975-01-01

    This paper evaluates the spatial characteristics of dynamic properties, especially soil-structure interaction behaviour, of the BWR nuclear building by experimental and analytical studies. It is well known that the damping effects in soil-structure interaction are remarkable on the building with short periods by the dissipation of vibrational energy to the soil. The authors have previously reported an analytical method for estimating the damping effects the properties of which are characterized as follows: 1) The complex damping is used, because the so-called structural damping may be more suitable for estimating the damping effects of an elastic structure. 2) H. Tajimi's theory is used for estimating the dynamical soil-foundation stiffness with the dissipation of vibrational energy on the elastic half-space soil. In this paper, an approximate explanation is presented in regard to the more developmental mathematical method for estimating the damping effects than the above-mentioned previous method, which is 'Modes Superposition Method for Multi-Degrees of Freedom System' with the constant complex stiffness showing the structural damping effects and the dynamical soil-foundation stiffness approximated by the linear or quadratic functions of the eigenvalues. An approximate explanation is presented in regard to the experimental results of the No. 1 reactor building (BWR) of Hamaoka Nuclear Power Station, The Chubu Electric Power Co., Ltd. (Auth.)

  3. An assessment of BWR [boiling water reactor] Mark-II containment challenges, failure modes, and potential improvements in performance

    International Nuclear Information System (INIS)

    Kelly, D.L.; Jones, K.R.; Dallman, R.J.; Wagner, K.C.

    1990-07-01

    This report assesses challenges to BWR Mark II containment integrity that could potentially arise from severe accidents. Also assessed are some potential improvements that could prevent core damage or containment failure, or could mitigate the consequences of such failure by reducing the release of fission products to the environment. These challenges and improvements are analyzed via a limited quantitative risk/benefit analysis of a generic BWR/4 reactor with Mark II containment. Point estimate frequencies of the dominant core damage sequences are obtained and simple containment event trees are constructed to evaluate the response of the containment to these severe accident sequences. The resulting containment release modes are then binned into source term release categories, which provide inputs to the consequence analysis. The output of the consequences analysis is used to construct an overall base case risk profile. Potential improvements and sensitivities are evaluated by modifying the event tree spilt fractions, thus generating a revised risk profile. Several important sensitivity cases are examined to evaluate the impact of phenomenological uncertainties on the final results. 75 refs., 25 figs., 65 tabs

  4. Application of the MOVE algorithm for the identification of reduced order models of a core of a BWR type reactor

    International Nuclear Information System (INIS)

    Victoria R, M.A.; Morales S, J.B.

    2005-01-01

    Presently work is applied the modified algorithm of the ellipsoid of optimal volume (MOVE) to a reduced order model of 5 differential equations of the core of a boiling water reactor (BWR) with the purpose of estimating the parameters that model the dynamics. The viability is analyzed of carrying out an analysis that calculates the global dynamic parameters that determine the stability of the system and the uncertainty of the estimate. The modified algorithm of the ellipsoid of optimal volume (MOVE), is a method applied to the parametric identification of systems, in particular to the estimate of groups of parameters (PSE for their initials in English). It is looked for to obtain the ellipsoid of smaller volume that guarantees to contain the real value of the parameters of the model. The PSE MOVE is a recursive identification method that can manage the sign of noise and to ponder it, the ellipsoid represents an advantage due to its easy mathematical handling in the computer, the results that surrender are very useful for the design of Robust Control since to smaller volume of the ellipsoid, better is in general the performance of the system to control. The comparison with other methods presented in the literature to estimate the reason of decline (DR) of a BWR is presented. (Author)

  5. 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)

  6. 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.)

  7. Summary of the OECD/NRC Boiling Water Reactor Turbine Trip Benchmark - Fourth Workshop (BWR-TT4)

    International Nuclear Information System (INIS)

    2002-01-01

    The reference problem chosen for simulation in a BWR is a Turbine Trip transient, which begins with a sudden Turbine Stop Valve (TSV) closure. The pressure oscillation generated in the main steam piping propagates with relatively little attenuation into the reactor core. The induced core pressure oscillation results in dramatic changes of the core void distribution and fluid flow. The magnitude of the neutron flux transient taking place in the BWR core is strongly affected by the initial rate of pressure rise caused by pressure oscillation and has a strong spatial variation. The correct simulation of the power response to the pressure pulse and subsequent void collapse requires a 3-D core modeling supplemented by 1-D simulation of the remainder of the reactor coolant system. A BWR TT benchmark exercise, based on a well-defined problem with complete set of input specifications and reference experimental data, has been proposed for qualification of the coupled 3-D neutron kinetics/thermal-hydraulic system transient codes. Since this kind of transient is a dynamically complex event with reactor variables changing very rapidly, it constitutes a good benchmark problem to test the coupled codes on both levels: neutronics/thermal-hydraulic coupling and core/plant system coupling. Subsequently, the objectives of the proposed benchmark are: comprehensive feedback testing and examination of the capability of coupled codes to analyze complex transients with coupled core/plant interactions by comparison with actual experimental data. The benchmark consists of three separate exercises: Exercise 1 - Power vs. Time Plant System Simulation with Fixed Axial Power Profile Table (Obtained from Experimental Data). Exercise 2 - Coupled 3-D Kinetics/Core Thermal-Hydraulic BC Model and/or 1-D Kinetics Plant System Simulation. Exercise 3 - Best-Estimate Coupled 3-D Core/Thermal-Hydraulic System Modeling. The purpose of this fourth workshop was to present and discuss final results of

  8. Modelling of the dynamics of the vessel and circuits of recirculation of a BWR type nucleo electric as part of the SUN-RAH university simulator; Modelado de la dinamica de la vasija y circuitos de recirculacion de una nucleoelectrica tipo BWR como parte del simulador universitario SUN-RAH

    Energy Technology Data Exchange (ETDEWEB)

    Sanchez S, R A [DEPFI, Campus Morelos, en IMTA, Jiutepec, Morelos (Mexico)

    2003-07-01

    In the present project, the development of a model for the dynamics of the process of energy transport generated in the nuclear fuel until the main steam lines of a nucleo electric central with BWR type nuclear reactor, using mathematical models of reduced order is presented. These models present the main characteristics of the reactor vessel and of the recirculation system, defined by the main phenomena that intervene in those physical processes. Likewise, the objective of the general project of the one University student nucleo electric simulator with Boiling Water Reactor (SUN-RAH) for later on to establish the modeling equations for each part of the nuclear reactor as well as of the load pursuit system. Also, its were described the graphic interfaces implemented in an three layers architecture in which the different measuring variables are presented in the monitor. It fits signalize that the advantage presented by the University student nucleo electric simulator is the possibility to carry out changes in the magnitudes of those different variables that intervene in the physical processes made in the one reactor and in the recirculation system in execution time of the same one. Of same way, the creation of a graphic intuitive interface, friendly, and designed with the same technology with the one that the video games are programmed in the present time. Besides all the above mentioned, the pending goals inside of the project are exposed, as well as the developments in construction process or conceptualized to be included in future versions of the simulator. Finally its are thinking about possible scenarios of applications of SUN-RAH, as well as their reaches. (Author)

  9. Siemens Nuclear Power Corporation methods development for BWR/PWR reactor licensing

    International Nuclear Information System (INIS)

    Pruitt, D.W.

    1992-01-01

    This presentation addresses the Siemens Nuclear Power Corporation (SNP) perspective on the primary forces driving methods development in the nuclear industry. These forces are fuel design, computational environment and industry requirement evolution. The first segment of the discussion presents the SNP experience base. SNP develops, manufactures and licenses both BWR and PWR reload fuel. A review of this experience base highlights the accelerating rate at which new fuel designs are being introduced into the nuclear industry. The application of advanced BWR lattice geometries provides an example of fuel design trends. The second aspect of the presentation is the rapid evolution of the computing environment. The final subject in the presentation is the impact of industry requirements on code or methods development

  10. Computing radiation dose to reactor pressure vessel and internals

    International Nuclear Information System (INIS)

    1996-01-01

    Within the next twenty years many of the nuclear reactors currently in service will reach their design lifetime. One of the key factors affecting decisions on license extensions will be the ability to confidently predict the integrity of the reactor pressure vessel and core structural components which have been subjected to many years of cumulative radiation exposure. This report gives an overview of the most recent scientific literature and current methodologies for computational dosimetry in the OECD/NEA Member countries. Discussion is extended to consider some related issues of materials science, such as the metals, and limitations of the models in current use. Proposals are made for further work. (author)

  11. 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

  12. MCTP, a code for the thermo-mechanical analysis of a fuel rod of BWR type reactors (Neutron part); MCTP, un codigo para el analisis termo-mecanico de una barra combustible de reactores tipo BWR (Parte Neutronica)

    Energy Technology Data Exchange (ETDEWEB)

    Hernandez L, H; Ortiz V, J [ININ, 52045 Ocoyoacac, Estado de Mexico (Mexico)

    2003-07-01

    In the National Institute of Nuclear Research of Mexico a code for the thermo-mechanical analysis of the fuel rods of the BWR type reactors of the Nucleo electric Central of Laguna Verde is developed. The code solves the diffusion equation in cylindrical coordinates with several energy groups. The code, likewise, calculates the temperature distribution and power distribution in those fuel rods. The code is denominated Multi groups With Temperatures and Power (MCTP). In the code, the energy with which the fission neutrons are emitted it is divided in six groups. They are also considered the produced perturbations by the changes in the temperatures of the materials that constitute the fuel rods, the content of fission products, the uranium consumption and in its case the gadolinium, as well as the plutonium production. In this work there are present preliminary results obtained with the code, using data of operation of the Nucleo electric Central of Laguna Verde. (Author)

  13. 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

  14. 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.)

  15. 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.

  16. 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)

  17. 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.

  18. 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.)

  19. 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

  20. Prestressed concrete reactor vessels: review of design and failure criteria

    International Nuclear Information System (INIS)

    Endebrock, E.G.

    1975-03-01

    The design and failure criteria of prestressed concrete reactor vessels (PCRVs) are reviewed along with the analysis methods. The mechanical properties of concrete under multiaxial stresses are not adequately quantified or described to permit an accurate analysis of a PCRV. Structural analysis of PCRVs almost universally utilizes a finite element which encounters difficulties in numerical solution of the governing equations and in treatment of fractured elements. (U.S.)

  1. Re-examining reactor vessel embrittlement at Chooz A

    International Nuclear Information System (INIS)

    Guilleret, J.-C.

    1988-01-01

    The Chooz A PWR experienced an extended shutdown in 1987/88 following indications that the reactor vessel was embrittling more rapidly than expected. Discrepancies between the expected rate and estimates of the actual rate were not easily explained. The huge body of work done since then to establish safety margins and support restart of the plant should provide a model for the owners of other older PWRs grappling with the embrittlement issue. (author)

  2. TMI-2 reactor vessel and balance of plant status

    International Nuclear Information System (INIS)

    Kuehn, G.A.

    1990-01-01

    In the fall of 1988 a corporate decision was made which concentrated effort on support of reactor vessel defueling and minimized activity in balance-of-plant areas. The auxiliary and fuel handling building are in a safe/stable state but final preparations for monitored storage won't be pursued until defueling and fuel shipping are complete. In addition to dispositioning fuel, the project is actively preparing for disposal of the Accident Generated Water (2.3 million gallons) by evaporation

  3. 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

  4. 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.)

  5. Decomissioning of nuclear reactors - methods for calculation of radionuclide inventories in contaminated BWR systems

    International Nuclear Information System (INIS)

    Lundgren, K.

    1991-01-01

    The purpose of the study has been to develop and demonstrate calculation models for the prediction of radionuclide inventories in contaminated systems in the Nordic BWRs at the time of decommissioning. Oskarshamn 2 was selected as reference reactor for the study. The study is divided in radionuclide inventories of activated corrosion products, and inventories of fission products and actinides from leaking fuel. The study is restricted to contamination outside the reactor pressure vessel. Inventories of activated corrosion products on primary system surfaces were predicted with the ABB Atom computer code BKM-CRUD. The calculations were performed for an extended operation time up to year 2010 for the nuclides Co60, Co58, Zn65 and Mn54. A special set of calculations were also made covering the non-standard nuclides Fe55, Ni59, No63 and Mo93. ABB Atom has carried out a comprehensive program on shutdown dose rates and activity measurements in delivered BWRs. The resulting data base have been used in the study to derive conversion factors for evaluation of contamination levels in secondary systems from the BKM-CRUD results for primary systems. Fission products and actinides were treated by defining two different fuel leakage scenarios. The first one corresponds to a rather stable situation, with an average leakage rate of 1 'standard pin hole' (i.e. 2 MBq/s Xe133). The second scenario means more severe fuel leakage every 10 years (200 MBq/s Xe133). The radioactive inventories in different part of the plant were calculated by combining the surface areas with the calculated contamination levels in the different systems. The uncertainty in total activity inventory has been estimated to not exceed a factor of 2. The uncertainty in total activity inventory has been to not exceed a factor of 2. The estimated inventories in this study has been compared to the results from other studies, and a reasonable agreement was achieved

  6. Assessment and management of ageing of major nuclear power plant components important to safety: BWR pressure vessels

    International Nuclear Information System (INIS)

    2005-10-01

    . The guidance reports are directed at technical experts from NPPs and from regulatory, plant design, manufacturing and technical support organizations dealing with specific plant components addressed in the reports. This report addresses the reactor pressure vessel (RPV) in BWRs. Maintaining the structural integrity of this RPV throughout NPP service life in spite of several ageing mechanisms is essential for plant safety

  7. 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)

  8. Assessment of two BWR accident management strategies

    International Nuclear Information System (INIS)

    Hodge, S.A.; Petek, M.

    1991-01-01

    A recently completed Oak Ridge effort proposes two management strategies for mitigation of the events that might occur in-vessel after the onset of significant core damage in a BWR severe accident. While the probability of such an accident is low, there may be effective yet inexpensive mitigation measures that could be implemented employing the existing plant equipment and requiring only additions to the plant emergency procedures. In this spirit, accident management strategies have been proposed for use of a borated solution for reactor vessel refill should control blade damage occur during a period of temporary core dryout and for containment flooding to maintain the core debris within the reactor vessel if injection systems cannot be restored. The proposed strategy for poisoning of the water used for vessel reflood should injection systems be restored after control blade damage has occurred has great promise, using only the existing plant equipment but employing a different chemical form for the boron poison. The dominant BWR severe accident sequence is Station Blackout and without means for mechanical stirring or heating of the storage tank, the question of being able to form the poisoned solution under accident conditions becomes of supreme importance. On the other hand, the proposed strategy for drywell flooding to cool the reactor vessel bottom head and prevent the core and structure debris from escaping to the drywell holds less promise. This strategy does, however, have potential for future plant designs in which passive methods might be employed to completely submerge the reactor vessel under severe accident conditions without the need for containment venting

  9. 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

  10. MCTP, a code for the thermo-mechanical analysis of a fuel rod of BWR type reactors (Neutron part)

    International Nuclear Information System (INIS)

    Hernandez L, H.; Ortiz V, J.

    2003-01-01

    In the National Institute of Nuclear Research of Mexico a code for the thermo-mechanical analysis of the fuel rods of the BWR type reactors of the Nucleo electric Central of Laguna Verde is developed. The code solves the diffusion equation in cylindrical coordinates with several energy groups. The code, likewise, calculates the temperature distribution and power distribution in those fuel rods. The code is denominated Multi groups With Temperatures and Power (MCTP). In the code, the energy with which the fission neutrons are emitted it is divided in six groups. They are also considered the produced perturbations by the changes in the temperatures of the materials that constitute the fuel rods, the content of fission products, the uranium consumption and in its case the gadolinium, as well as the plutonium production. In this work there are present preliminary results obtained with the code, using data of operation of the Nucleo electric Central of Laguna Verde. (Author)

  11. Experimental and analytical studies for a BWR nuclear reactor building evaluation of soil-structure interaction behavior

    International Nuclear Information System (INIS)

    Mizuno, N.; Tsushima, Y.

    1975-01-01

    The purpose of this paper is to evaluate the spatial characteristics of dynamic properties, especially soil-structure interaction behavior, or the BWR nuclear reactor building by experimental and analytical studies. An analytical method (SMIRT-1 Paper K 2/4) for estimating the damping effects is reported. The complex damping is used, because the so-called structural damping may be more suitable for estimating the damping effects of an elastic structure. H. Tajimi's theory is used for estimating the dynamical soil-foundation stiffness with the dissipation of vibrational energy on the elastic half-space soil. An approximate explanation is presented in regard to the more developmental mathematical method for estimating the damping effects than the above-mentioned previous method, which is 'Modes Superposition Method for Multi-Degrees of Freedom System' with the constant complex stiffness showing the structural damping effects and the dynamical soil-foundation stiffness approximated by the linear or quadratic functions of the eigenvalues. Next, an approximate explanation is presented in regard to the experimental results of the No.1 reactor building (BWR) of Hamaoka Nuclear Power Station, The Chubu Electric Power Co., Ltd. The regression analyses of the experimental resonance curves by one degree system show that the critical damping ratio is larger than the 0.10 used in the design for the fundamental natural period. It is attempted to simulate the experimental results by the above-mentioned method. The simulated model is a fourty-eight degrees of freedom spring mass system because of the eight masses for the eight floors including the base foundation and the six degrees of freedom for a mass

  12. 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.

  13. 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)

  14. Stress analysis of blanket vessel for JAERI experimental fusion reactor

    International Nuclear Information System (INIS)

    Sako, K.; Minato, A.

    1979-01-01

    A blanket structure of JAERI Experimental Fusion Reactor (JXFR) consists of about 2,300 blanket cells with round cornered rectangular cross sections (twelve slightly different shapes) and is placed in a vacuum vessel. Each blanket vessel is a double-walled thin-shell structure made of Type 316 stainless steel with a spherical domed surface at the plasma side. Ribs for coolant channel are provided between inner and outer walls. The blanket cell contains Li 2 O pebbles and blocks for tritium breeding and stainless steel blocks for neutron reflection. A coolant is helium gas at 10 kgf/cm 2 (0.98 MPa) and its inlet and outlet temperatures are 300 0 C and 500 0 C. The maxima of heat flux and nuclear heating rate at the first wall are 12 W/cm 2 and 2 W/cc. A design philosophy of the blanket structure is based on high tritium breeding ratio and more effective shielding performance. The thin-shell vessel with a rectangular cross section satisfies the design philosophy. We have designed the blanket structure so that the adjacent vessels are mutually supporting in order to decrease the large deformation and stress due to internal pressure in case of the thin-shell vessel. (orig.)

  15. Vessel core seismic interaction for a fast reactor

    International Nuclear Information System (INIS)

    Martelli, A.; Maresca, G.

    1984-01-01

    This report deals with the analysis carried out in collaboration between ENEA and NIRA for optimizing the iterative procedure applied for the evaluation of the effects of the vessel core dynamic interaction for a fast reactor in the case of a earthquake. In fact, as shown in a previous report the convergence of such procedure was very slow for the design solution adopted for the PEC reactor, i.e. with a core restraint plate located close to the top of the core elements. This study, although performed making use of preliminary data (the same of the cited previous report) demonstrates that the convergence is fast if a suitable linear core model is applied in the first iteration linear calculations carried out by NIRA, with an intermediate stiffness with respect to those corresponding to the two limit models previously assumed and increased damping coefficients. Thus, the optimized iterative procedures is now applied in the PEC reactor block seismic verification analysis

  16. 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

  17. 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

  18. 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

  19. 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)

  20. 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.

  1. 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

  2. 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.

  3. 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.

  4. 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

  5. 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.

  6. 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

  7. Comparison of results for burning with BWR reactors CASMO and SCALE 6.2 (TRITON / NEWT); Comparacion de los resultados de quemado para reactores BWR con CASMO y SCALE 6.2 (TRITON/NEWT)

    Energy Technology Data Exchange (ETDEWEB)

    Mesado, C.; Miro, R.; Barrachina, T.; Verdu, G.

    2014-07-01

    In this paper we compare the results from two codes burned, CASMO and SCALE 6.2 (TRITON). To do this, is simulated all segments corresponding to a boiling water reactor (BWR) using both codes. In addition, to account for different working points, simulations changing the instantaneous variables, these are repeated: void fractions (6 points), fuel temperature (6 points) and control rods (two points), with a total of 72 possible combinations of different instantaneous variables for each segment. After all simulations are completed for each segment, we can reorder the obtained cross sections, as SCALE CASMO both, to create a library of compositions nemtab format. This format is accepted by the neutronic code of nodal diffusion, PARCS v2.7. Finally compares the results obtained with PARCS and with the SIMULATE3 -SIMTAB methodology to level of full reactor. Also, we have made use of the KENO-VI and MCDANCOFF modules belonging to SCALE. The first is a Monte Carlo transport code with which you can validate the value of the multiplier, the second has been used to obtain values of Dancoff factor and increase the accuracy of model SCALE. (Author)

  8. Analysis of Reactor Vessel Lower Head Penetration Tube Failure

    International Nuclear Information System (INIS)

    Stempniewicz, Marek

    1999-01-01

    This paper presents results of two studies, performed to investigate the behavior of the reactor vessel penetration tubes in case of relocation of molten material into the tubes. The first study is on the CORVIS drain line experiment 03/1. Results of pre-test calculations are presented, and compared to the later obtained experimental data. The timing of the drain line melting and the velocity of the debris flowing inside the drain line were predicted correctly, but the penetration depth was clearly underestimated. If the calculations are done using different correlation for the melt-to-wall convective heat transfer, the results are closer to the experiment. It cannot however be concluded that the alternative correlation is more appropriate until other uncertainties are clarified. The second study presents calculations performed for GKN Dodewaard CRD, instrument tubes and drain line. Calculations were performed to estimate whether the tubes have a chance to withstand the first attack of the melt and thus postpone vessel failure until the water in the lower plenum evaporates. Calculations were performed assuming that the melt can move into the tubes without any resistance, e.g. presence of water in the tubes was not taken into account. The results indicate that the critical penetration of the GKN vessel, which is most likely to fail, is the drain line. Results also indicate that external flooding should prevent early tube failure, at least in case of low vessel pressure. (author)

  9. Automatic refueling platform and CRD remote handling device for BWR plant

    International Nuclear Information System (INIS)

    Kato, Hiroaki; Takagi, Kaoru

    1978-01-01

    In BWR plants, machines for replacing fuel assemblies and control rod drives are usually operated directly by personnel. An automatic refueling platform and a CRD remote handling device aiming at radiation exposure reduction and handling perfectness are described, which are already used in BWR plants. Automation of the former is achieved in transporting fuel assemblies between a reactor pressure vessel and a fuel storage pool, shuffling fuel assemblies in a reactor core and moving fuel assemblies in a fuel storage pool. In the latter, replacement of CRDs is nearly all performed remotely. (Mori, K.)

  10. ZZ BWRSB-RINGHALS1, Stability Benchmark Data from BWR RINGHALS-1

    International Nuclear Information System (INIS)

    2002-01-01

    Description of program or function: The purpose of this benchmark is to enable code developers to test their codes and also to validate the predictive capability of their respective codes and models for BWR stability analysis. Emphasis is put on the modelling of flow dynamics of the reactor core and in-vessel flow loop wit detailed neutronic and thermodynamic feedback. The secondary systems as well as the control and production systems will be neglected. Data provided comes from measurements in beginning of cycle (BOC) 14, 15, 16 and 17 and middle of cycle (MOC) 16 in the Swedish BWR reactor Ringhals 1. For these measurements complete data sets are given

  11. Ex-vessel coolability and energetics of steam explosions in nordic light water reactors - EXCOOLSE project report 2004

    Energy Technology Data Exchange (ETDEWEB)

    Park, H.S.; Nayak, A.K.; Hansson, R.C.; Sehgal, B.R. [Royal Inst. of Technology, Div. of Nuclear Power Safety (Sweden)

    2005-10-01

    Beyond-the-design-basis accidents, i.e. severe accidents, involve melting of the nuclear reactor core and release of radioactivity. Intensive research has been performed for years to evaluate the consequence of the postulated severe accidents. Severe accidents posed, to the reactor researchers, a most interesting and most difficult set of phenomena to understand, and to predict the consequences, for the various scenarios that could be contemplated. The complexity of the interactions, occurring at such high temperatures ({approx} 2500 deg. C), between different materials, which are changing phases and undergoing chemical reactions, is simply indescribable with the accuracy that one may desire. Thus, it is a wise approach to pursue research on SA phenomena until the remaining uncertainty in the predicted consequence, or the residual risk, can be tolerated. In the PRE-DELI-MELT project at NKS, several critical issues on the core melt loadings in the BWR and PWR reactor containments were identified. Many of Nordic nuclear power plants, particularly in boiling water reactors, adopted the Severe Accident Management Strategy (SAMS) which employed the deep subcooled water pool in lower dry-well. The success of this SAMS largely depends on the issues of steam explosions and formation of debris bed and its coolability. From the suggestions of the PRE-DELI-MELT project, a series of research plan was proposed to investigate the remaining issues specifically on the ex-vessel coolability of corium during severe accidents; (a) ex-vessel coolability of the melt or particulate debris, and (b) energetics and debris characteristics of fuel-coolant interactions endangering the integrity of the reactor containments. (au)

  12. Ex-vessel coolability and energetics of steam explosions in nordic light water reactors - EXCOOLSE project report 2004

    International Nuclear Information System (INIS)

    Park, H.S.; Nayak, A.K.; Hansson, R.C.; Sehgal, B.R.

    2005-10-01

    Beyond-the-design-basis accidents, i.e. severe accidents, involve melting of the nuclear reactor core and release of radioactivity. Intensive research has been performed for years to evaluate the consequence of the postulated severe accidents. Severe accidents posed, to the reactor researchers, a most interesting and most difficult set of phenomena to understand, and to predict the consequences, for the various scenarios that could be contemplated. The complexity of the interactions, occurring at such high temperatures (∼ 2500 deg. C), between different materials, which are changing phases and undergoing chemical reactions, is simply indescribable with the accuracy that one may desire. Thus, it is a wise approach to pursue research on SA phenomena until the remaining uncertainty in the predicted consequence, or the residual risk, can be tolerated. In the PRE-DELI-MELT project at NKS, several critical issues on the core melt loadings in the BWR and PWR reactor containments were identified. Many of Nordic nuclear power plants, particularly in boiling water reactors, adopted the Severe Accident Management Strategy (SAMS) which employed the deep subcooled water pool in lower dry-well. The success of this SAMS largely depends on the issues of steam explosions and formation of debris bed and its coolability. From the suggestions of the PRE-DELI-MELT project, a series of research plan was proposed to investigate the remaining issues specifically on the ex-vessel coolability of corium during severe accidents; (a) ex-vessel coolability of the melt or particulate debris, and (b) energetics and debris characteristics of fuel-coolant interactions endangering the integrity of the reactor containments. (au)

  13. 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)

  14. Prototype fast reactor steam generator unit pressure vessel repairs

    International Nuclear Information System (INIS)

    Daniels, B.D.; Green, D.; Henderson, J.D.C.

    1993-01-01

    The prototype fast reactor at Dounreay has experienced a number of unscheduled shutdowns due to leaking reheater and superheater shell welds. There was a need to determine the cracking mechanism and to design a general repair technique simultaneously. Detailed investigations revealed that the crack locations correlated with the positions of rectification welds made at the time of vessel manufacture. A creep crack growth mechanism was identified; this requires through wall residual stress for through cracks to develop. A repair technique has been devised and successfully applied to the sites of a number of leaks. (author)

  15. Advanced ultrasonic and eddy current examinations of the reactor vessel

    International Nuclear Information System (INIS)

    Cvitanovic, M.; Zado, V.

    1996-01-01

    In order to improve safety and reliability of nuclear power plant components, the existing examination methods are permanently developed as well as the new methods of examination are implemented. For the same reason, beside referent requirements, complementary NDE methods are utilized. Some examination methods techniques are not required to be used by referent safety codes and standards but they are frequently practiced as additional prevention to the component failure. This article presents the state of the art methods and techniques currently applied for examination of the reactor vessel base material, clad and weld materials. (author)

  16. Reactor pressure vessel embrittlement: Insights from neural network modelling

    Science.gov (United States)

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

    2018-04-01

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

  17. Annealing of the BR3 reactor pressure vessel

    International Nuclear Information System (INIS)

    Fabry, A.; Motte, F.; Stiennon, G.; Debrue, J.; Gubel, P.; Van de Velde, J.; Minsart, G.; Van Asbroeck, P.

    1985-01-01

    The pressure vessel of the Belgian BR-3 plant, a small (11 MWe) PWR presently used for fuel testing programs and operated since 1962, was annealed during March, 1984. The anneal was performed under wet conditions for 168 hours at 650 0 F with core removal and within plant design margins justification for the anneal, summary of plant characteristics, description of materials sampling, summary of reactor physics and dosimetry, development of embrittlement trend curves, hypothesized pressurized and overcooling thermal shock accidents, and conclusions are provided in detail

  18. Automated ultrasonic shop inspection of reactor pressure vessel forgings

    International Nuclear Information System (INIS)

    Farley, J.M.; Dikstra, B.J.; Hanstock, D.J.; Pople, C.H.

    1986-01-01

    Automated ultrasonic shop inspection utilizing a computer-controlled system is being applied to each of the forgings for the reactor pressure vessel of the proposed Sizewell B PWR power station. Procedures which utilize a combination of high sensitivity shear wave pulse echo, 0 degrees and 70 degrees angled longitudinal waves, tandem and through-thickness arrays have been developed to provide comprehensive coverage and an overall reliability of inspection comparable to the best achieved in UKAEA defect detection trials and in PISC II. This paper describes the ultrasonic techniques, the automated system (its design, commissioning and testing), validation and the progress of the inspections

  19. 3 D flow computations under a reactor vessel closure head

    International Nuclear Information System (INIS)

    Daubert, O.; Bonnin, O.; Hofmann, F.; Hecker, M.

    1995-12-01

    The flow under a vessel cover of a pressurised water reactor is investigated by using several computations and a physical model. The case presented here is turbulent, isothermal and incompressible. Computations are made with N3S code using a k-epsilon model. Comparisons between numerical and experimental results are on the whole satisfying. Some local improvements are expected either with more sophisticated turbulence models or with mesh refinements automatically computed by using the adaptive meshing technique which has been just implemented in N3S for 3D cases. (authors). 6 refs., 7 figs

  20. Pressure thermal shock analysis for nuclear reactor pressure vessel

    International Nuclear Information System (INIS)

    Galik, G.; Kutis, V.; Jakubec, J.; Paulech, J.; Murin, J.

    2015-01-01

    The appearance of structural weaknesses within the reactor pressure vessel or its structural failure caused by crack formation during pressure thermal shock processes pose as a severe environmental hazard. Coolant mixing during ECC cold water injection was simulated in a detailed CFD analysis. The temperature distribution acting on the pipe wall internal surface was calculated. Although, the results show the formation of high temperature differences and intense gradients, an additional structural analysis is required to determine the possibility of structural damage from PTS. Such an analysis will be the subject of follow-up research. (authors)

  1. Optimization of reactor pressure vessel internals segmentation in Korea

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Byung-Sik [Dankook Univ., Chungnam (Korea, Republic of). Dept. of Nuclear Engineering

    2017-11-15

    One of the most challenging tasks during plant decommissioning is the removal of highly radioactive internal components from the reactor pressure vessel (RPV). For RPV internals dismantling, it is essential that all activities are thoroughly planned and discussed in the early stage of the decommissioning project. One of the key activities in the detailed planning is to prepare the segmentation and packaging plan that describes the sequential steps required to segment, separate, and package each individual component of RPV, based on an activation analysis and component characterization study.

  2. Needs for evaluated covariance data for reactor pressure vessel dosimetry

    International Nuclear Information System (INIS)

    Maerker, R.E.; Broadhead, B.L.; Wagschal, J.J.

    1992-01-01

    This report discusses new methodology for quantifying and then reducing uncertainties in the calculated pressure vessel fluences of a pressurized water reactor (PWR). The technique involves combining the integral results of the calculated and measured PWR surveillance dosimetry activities with the differential data used in the calculations, along with covariances of all the quantities, into a generalized linear least-squares adjustment procedure. Based on analysis of both PWRs and test reactor benchmarks, substantial evidence now exists to support the conclusion that, of all the nuclear as well as non-nuclear differential data considered, ENDF/B-VI values of the total inelastic iron cross sections and their covariances are the most important data controlling the outcome of the adjustment procedure. Predicted adjustments in these cross sections provided the stimulus for new measurements, the results of which impacted the ENDF/B-VI evaluation of iron 56

  3. Crystal Plasticity Model of Reactor Pressure Vessel Embrittlement in GRIZZLY

    Energy Technology Data Exchange (ETDEWEB)

    Chakraborty, Pritam [Idaho National Laboratory (INL), Idaho Falls, ID (United States); Biner, Suleyman Bulent [Idaho National Laboratory (INL), Idaho Falls, ID (United States); Zhang, Yongfeng [Idaho National Laboratory (INL), Idaho Falls, ID (United States); Spencer, Benjamin Whiting [Idaho National Laboratory (INL), Idaho Falls, ID (United States)

    2015-07-01

    The integrity of reactor pressure vessels (RPVs) is of utmost importance to ensure safe operation of nuclear reactors under extended lifetime. Microstructure-scale models at various length and time scales, coupled concurrently or through homogenization methods, can play a crucial role in understanding and quantifying irradiation-induced defect production, growth and their influence on mechanical behavior of RPV steels. A multi-scale approach, involving atomistic, meso- and engineering-scale models, is currently being pursued within the GRIZZLY project to understand and quantify irradiation-induced embrittlement of RPV steels. Within this framework, a dislocation-density based crystal plasticity model has been developed in GRIZZLY that captures the effect of irradiation-induced defects on the flow stress behavior and is presented in this report. The present formulation accounts for the interaction between self-interstitial loops and matrix dislocations. The model predictions have been validated with experiments and dislocation dynamics simulation.

  4. Reactor vessel and core two-phase flow ultrasonic densitometer

    International Nuclear Information System (INIS)

    Arave, A.E.

    1979-01-01

    A local ultrasonic density (LUD) detector has been developed by EG and G Idaho, Inc., at the Idaho National Engineering Laboratory (INEL) for the Loss-of-Fluid Test (LOFT) reactor vessel and core two-phase flow density measurements. The principle of operating the sensor is the change in propagation time of a torsional ultrasonic wave in a metal transmission line as a function of the density of the surrounding media. A theoretical physics model is presented which represents the total propagation time as a function of the sensor modulus of elasticity and polar moment of inertia. Separate effects tests and two-phase flow tests have been conducted to characterize the detector. Tests show the detector can perform in a 343 0 C pressurized water reactor environment and measure the average density of the media surrounding the sensor

  5. Shear strength of end slabs of prestressed concrete reactor vessels

    International Nuclear Information System (INIS)

    Cheung, K.C.; Gotschall, H.L.; Liu, T.C.

    1975-01-01

    Prestressed concrete reactor vessels (PCRV's) have been adopted for primary containments in most large high-temperature gas-cooled reactor installations. The most common configuration for PCRVs is a right-vertical cylinder with thick end slabs. In order to assess the integrity of a PCRV it is necessary to predict the ultimate strength of the end slabs. The complexity of the basic mechanism of shear failure in the PCRV end slabs has thus far prohibited the development of a completely analytical solution. However, many experimental investigations of PCRV end slabs have been conducted over the past decade. This information makes it possible to establish empirical formulae for the ultimate strength of PCRV end slabs. The basis and development of an empirical shear-flexure interaction expression is presented. (Auth.)

  6. Crystal Plasticity Model of Reactor Pressure Vessel Embrittlement in GRIZZLY

    International Nuclear Information System (INIS)

    Chakraborty, Pritam; Biner, Suleyman Bulent; Zhang, Yongfeng; Spencer, Benjamin Whiting

    2015-01-01

    The integrity of reactor pressure vessels (RPVs) is of utmost importance to ensure safe operation of nuclear reactors under extended lifetime. Microstructure-scale models at various length and time scales, coupled concurrently or through homogenization methods, can play a crucial role in understanding and quantifying irradiation-induced defect production, growth and their influence on mechanical behavior of RPV steels. A multi-scale approach, involving atomistic, meso- and engineering-scale models, is currently being pursued within the GRIZZLY project to understand and quantify irradiation-induced embrittlement of RPV steels. Within this framework, a dislocation-density based crystal plasticity model has been developed in GRIZZLY that captures the effect of irradiation-induced defects on the flow stress behavior and is presented in this report. The present formulation accounts for the interaction between self-interstitial loops and matrix dislocations. The model predictions have been validated with experiments and dislocation dynamics simulation.

  7. ASTM Standards for Reactor Dosimetry and Pressure Vessel Surveillance

    International Nuclear Information System (INIS)

    GRIFFIN, PATRICK J.

    1999-01-01

    The ASTM standards provide guidance and instruction on how to field and interpret reactor dosimetry. They provide a roadmap towards understanding the current ''state-of-the-art'' in reactor dosimetry, as reflected by the technical community. The consensus basis to the ASTM standards assures the user of an unbiased presentation of technical procedures and interpretations of the measurements. Some insight into the types of standards and the way in which they are organized can assist one in using them in an expeditious manner. Two example are presented to help orient new users to the breadth and interrelationship between the ASTM nuclear metrology standards. One example involves the testing of a new ''widget'' to verify the radiation hardness. The second example involves quantifying the radiation damage at a pressure vessel critical weld location through surveillance dosimetry and calculation

  8. 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

  9. 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

  10. 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.

  11. The modeling of core melting and in-vessel corium relocation in the APRIL code

    Energy Technology Data Exchange (ETDEWEB)

    Kim. S.W.; Podowski, M.Z.; Lahey, R.T. [Rensselaer Polytechnic Institute, Troy, NY (United States)] [and others

    1995-09-01

    This paper is concerned with the modeling of severe accident phenomena in boiling water reactors (BWR). New models of core melting and in-vessel corium debris relocation are presented, developed for implementation in the APRIL computer code. The results of model testing and validations are given, including comparisons against available experimental data and parametric/sensitivity studies. Also, the application of these models, as parts of the APRIL code, is presented to simulate accident progression in a typical BWR reactor.

  12. 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

  13. State of the art of second international exercise on benchmarks in BWR reactors

    International Nuclear Information System (INIS)

    Verdu, G.; Munoz-Cobo, J. L.; Palomo, M. J.; Escriva, A.; Ginestar, D.

    1998-01-01

    This is a second in series of Benchmarks based on data from operating Swedish BWRs. The first one concerned measurements made in cycles 14,15 16 and 17 at Ringhals 1 Nuclear Power Plant and addressed predictive power of analytical tools used in BWR stability analysis. Part of the data was disclosed only after participants had provided their results. This work has been published in the report: NEA/NSC/DOC(96)22, November 1996. In this report it was recognised that there is a need for better qualification of the applied noise analysis methods. A follow up Benchmark was thus proposed dedicated to the analysis of time series data and including the evaluation of both global and regional stability of Forsmarks 1 and 2 Nuclear Power Plant. In this second Benchmark have participated Forsmarks Kraftgrupp AB,NEA Nuclear Science Committee, CSN Consejo de Seguridad Nuclear and Department of Chemical and Nuclear Engineering of Polytechnic University of Valencia. (Author)

  14. Simulation of the aspersion system of the core low pressure (LPCS) for a boiling water reactor (BWR) based on RELAP

    International Nuclear Information System (INIS)

    Membrillo G, O. E.; Chavez M, C.

    2012-10-01

    The present work presents the modeling and simulation of the aspersion system to low pressure of reactor of the nuclear power plant of Laguna Verde using the nuclear code RELAP/SCDAP. The objective of the emergency systems inside a nuclear reactor is the cooling of the core, nor caring the performance of any other emergency system in the case of an accident design base for coolant loss. To obtain a simulation of the system is necessary to have a model based on their main components, pipes, pumps, valves, etc. This article describes the model for the simulation of the main line and the test line for the HPCS. At the moment we have the simulation of the reactor vessel and their systems associated to the nuclear power plant of Laguna Verde, this work will allow to associate the emergency system model LPCS to the vessel model. The simulation of the vessel and the emergency systems will allow knowing the behavior of the reactor in the stage of the coolant loos, giving the possibility to analyze diverse scenarios. The general model will provide an auxiliary tool for the training in classroom and at distance in the operation of nuclear power plants. (Author)

  15. 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

  16. Construction of an external electrode for determination of electrochemical corrosion potential in normal operational conditions of an BWR type reactor for hot cells

    International Nuclear Information System (INIS)

    Aguilar T, J.A.; Rivera M, H.; Hernandez C, R.

    2001-01-01

    The behavior of the corrosion processes at high temperature requires of external devices that being capable to resist a temperature of 288 Centigrade and a pressure of 80 Kg/cm 2 , to give stable and reproducible results of some variable and resisting physically and chemically the radiation. The external electrode of Ag/AgCl fulfils all the requirements in the determination of the electrochemical corrosion potential under normal operational conditions of a BWR type reactor in hot cells. (Author)

  17. 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)

  18. 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

  19. 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)

  20. 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

  1. 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)

  2. Simplified compact containment BWR plant

    International Nuclear Information System (INIS)

    Heki, H.; Nakamaru, M.; Tsutagawa, M.; Hiraiwa, K.; Arai, K.; Hida, T.

    2004-01-01

    The reactor concept considered in this paper has a small power output, a compact containment and a simplified BWR configuration with comprehensive safety features. The Compact Containment Boiling Water Reactor (CCR), which is being developed with matured BWR technologies together with innovative systems/components, is expected to prove attractive in the world energy markets due to its flexibility in regard to both energy demands and site conditions, its high potential for reducing investment risk and its safety features facilitating public acceptance. The flexibility is achieved by CCR's small power output of 300 MWe class and capability of long operating cycle (refueling intervals). CCR is expected to be attractive from view point of investment due to its simplification/innovation in design such as natural circulation core cooling with the bottom located short core, internal upper entry control rod drives (CRDs) with ring-type dryers and simplified ECCS system with high pressure containment concept. The natural circulation core eliminates recirculation pumps and the maintenance of such pumps. The internal upper entry CRDs reduce the height of the reactor vessel (RPV) and consequently reduce the height of the primary containment vessel (PCV). The safety features mainly consist of large water inventory above the core without large penetration below the top of the core, passive cooling system by isolation condenser (IC), passive auto catalytic recombiner and in-vessel retention (IVR) capability. The large inventory increases the system response time in the case of design-base accidents, including loss of coolant accidents. The IC suppresses PCV pressure by steam condensation without any AC power. The recombiner decreases hydrogen concentration in the PCV in the case of a severe accident. Cooling the molten core inside the RPV if the core should be damaged by loss of core coolability could attain the IVR. The feasibility of CCR safety system has been confirmed by LOCA

  3. Remote controlled stud bolt handling device for reactor pressure vessel

    International Nuclear Information System (INIS)

    Shindo, Takenori; Shigehiro, Katsuya; Ito, Morio; Okada, Kenji

    1988-01-01

    In nuclear power stations, at the time of regular inspection, the works of opening and fixing the upper covers of reactor pressure vessels are carried out for inspecting the inside of reactor pressure vessels and exchanging fuel rods. These upper covers are fastened with many stud bolts, therefore, the works of opening and fixing require a large amount of labor, and are done under the restricted condition of wearing protective clothings and masks. Babcock Hitachi K.K. has completed the development of a remotely controlled automatic bolt tightenig device for this purpose, therefore, its outline is reported. The conventional method of these works and the problems in it are described. The design of the new device aimed at the parallel execution of cleaning screw threads, loosening and tightening nuts, and taking off and putting on nuts and washers, thus contributing to the shortening of regular inspection period, the reduction of the radiation exposure of workers, and the decrease of the number of workers. The function, reliability and endurance of the new device were confirmed by the verifying test using a device made for trial. The device is composed of a stand, a rail and four stations each with a cleaning unit, a stud tensioner and a nut handling unit. (K.I.)

  4. 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

  5. Microstructure and embrittlement of VVER 440 reactor pressure vessel steels

    International Nuclear Information System (INIS)

    Hennion, A.

    1999-03-01

    27 VVER 440 pressurised water reactors operate in former Soviet Union and in Eastern Europe. The pressure vessel, is made of Cr-Mo-V steel. It contains a circumferential arc weld in front of the nuclear core. This weld undergoes a high neutron flux and contains large amounts of copper and phosphorus, elements well known for their embrittlement potency under irradiation. The embrittlement kinetic of the steel is accelerated, reducing the lifetime of the reactor. In order to get informations on the microstructure and mechanical properties of these steels, base metals, HAZ, and weld metals have been characterized. The high amount of phosphorus in weld metals promotes the reverse temper embrittlement that occurs during post-weld heat treatment. The radiation damage structure has been identified by small angle neutron scattering, atomic probe, and transmission electron microscopy. Nanometer-sized clusters of solute atoms, rich in copper with almost the same characteristics as in western pressure vessels steels, and an evolution of the size distribution of vanadium carbides, which are present on dislocation structure, are observed. These defects disappear during post-irradiation tempering. As in western steels, the embrittlement is due to both hardening and reduction of interphase cohesion. The radiation damage specificity of VVER steels arises from their high amount of phosphorus and from their significant density of fine vanadium carbides. (author)

  6. 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)

  7. Material problems in accident analysis of prestressed concrete reactor vessels

    International Nuclear Information System (INIS)

    Bazant, Z.P.

    1977-01-01

    Due to their very high energy absorption capability, as well as their inherent safety advantages, prestressed concrete reactor vessels are presently being keenly studied as the basic barrier to contain hypothetical core disruptive accidents in a fast breeder reactor. One problem investigated is the nonlinear constitutive behavior and failure criteria for concrete. Previously, a comprehensive theory, called endochronic theory, has been shown to satisfy all basic currently known features of test data. Nevertheless uncertainty still exists with regard to non-proportional loading paths, for which good test data are lacking at present. An extension of the endochronic theory which correlates best with general experimental evidence and includes fracturing terms is given, and a comparison with vertex-type hardening in plasticity is made. A second problem which must be analysed in accident situations is the high temperature shock on the concrete walls (due to liquid sodium, up to 850 0 C). Refining a previous crude formulation, a rational model for calculating moisture and heat transfer and pore pressures in concrete subjected to thermal shock is presented. In conclusion, a new design concept, in which the concrete vessel is completely dehydrated and kept hot throughout its service life in order to substantially improve its response to thermal shock as well as liquid sodium contact, is described. (Auth.)

  8. Mitigation strategies of intergranular corrosion in systems of reactors of water boiling (BWR). Combined action of the chemistry of the hydrogen and the oxygen; Estrategias de mitigacion de la corrosion intergranular en sistemas de reactores de agua en ebullicion (BWR). Accion combinada de la quimica del hidrogeno y del oxigeno

    Energy Technology Data Exchange (ETDEWEB)

    Verdugo, M.

    2015-07-01

    Inter-Granular Stress Corrosion cracking (IGSCC) in austenitic stainless steel and in austenitic nickel-based alloys has been the subject of many studies the aim of which was to resolve one of the main problems faced by BWR nuclear power plants since the 1960s. This corrosion phenomenon is the result of the combined action of three factors: sensitization of the material, high local stresses and an aggressive medium. This paper deals with these factors separately and analyzes the oxidative chemistry of BWR reactors (aggressivity of the medium) as one the main causes if IGSCC. (Author)

  9. Nuclear reactors sited deep underground in steel containment vessels

    Energy Technology Data Exchange (ETDEWEB)

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

    2006-07-01

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

  10. Pressure test method for reactor pressure vessel in construction field

    International Nuclear Information System (INIS)

    Takeda, Masakado; Ushiroda, Koichi; Miyahara, Ryohei; Takano, Hiroshi; Matsuura, Tadashi; Sato, Keiya.

    1998-01-01

    Plant constitutional parts as targets of both of a primary pressure test and a secondary pressure test are disposed in communication with a reactor pressure vessel, and a pressure of the primary pressure test is applied to the targets of both tests, so that the primary pressure test and the second pressure test are conducted together. Since the number of pressure tests can be reduced to promote construction, and the number of workers can also be reduced. A pressure exceeding the maximum pressure upon use is applied to the pressure vessel after disposing the incore structures, to continuously conduct the primary pressure test and the secondary pressure test joined together and an incore flowing test while closing the upper lid of the pressure vessel as it is in the construction field. The number of opening/closing of the upper lid upon conducting every test can be reduced, and since the pressure resistance test is conducted after arranging circumference conditions for the incore flowing test, the tests can be conducted collectively also in view of time. (N.H.)

  11. Revisiting the reactor pressure vessel for long-time operation

    International Nuclear Information System (INIS)

    Lapena, J.; Serrano, M.; Diego, G. de; Hernandez Mayoral, M.

    2013-01-01

    The reactor pressure vessel (RPV) is one of the key components of nuclear power plants, especially for long time operation. It is a non-replaceable component, at least with current technology. the structural integrity of the vessel is evaluated within called monitoring programs where the degradation of the mechanical properties due to neutron irradiation is determined. From the first designs of the RPVs and monitoring programs in the years 60-70 currently still in force, there have been major advances in the understanding of radiation damage and methods of evaluation. Thus, it is recommended the use of forgings instead of plates in the construction of the RPVs in order to reduce the number of welds, more sensitive to neutron irradiation, and using starting materials with less content of impurities, particularly copper. To evaluate the embrittlement of RPVs the Master Curve methodology is currently used, through the testing of the charpy specimens from the surveillance capsules, to determine the fracture toughness. This article summarizes the last activities of CIEMAT into the European research projects LONGIIFE and PERFORM60, about the knowledge of radiation damage in materials with low copper content, traditionally considered less sensitive to irradiation, and the use of the Master Curve in advanced surveillance programs. The activities related to the problems associated with the use of large forging, such as the appearance of hydrogen flakes in the vessel of Doel 3, and its implications, are also presented. (Author)

  12. East/west steels for reactor pressure vessels

    International Nuclear Information System (INIS)

    Davies, M.; Kryukov, A.; Nikolaev, Y.; English, C.

    1997-01-01

    The report consist of three parts dealing with comparison of the irradiation behaviour of 'Eastern' and 'Western' steels, mechanisms of irradiation embrittlement and the role of compositional variations on the irradiation sensitivity of pressure vessels. Nickel, copper and phosphorus are the elements rendering the most essential influence on behaviour of pressure vessel steels under irradiation and subsequent thermal annealing. For WWER-440 reactor pressure vessel (RPV) steels in which nickel content does nor exceed 0.3% the main affecting factors are phosphorous and copper. For WWER-1000 RPV welds in which nickel content generally exceed 1.5% the role of nickel in radiation embrittlement is decisive. In 'Western' type steels main influencing elements are nickel and copper. The secondary role of phosphorus in radiation embrittlement of 'Western' steels is caused by lower relative content compared to 'Eastern' steels. The process of how copper, phosphorus and nickel contents affect the irradiation sensitivity of both types of steel seem to be similar. Some distinctions between the observed radiation effects is apparently caused by differences in the irradiation conditions and ratios of the contents of above mentioned elements in both types of steel. For 'Eastern' RPV steels the dependence of the recovery degree of irradiated steels due to postirradiation thermal annealing id obviously dependent on phosphorus contents and the influence of nickel contents on this process is detectable

  13. 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)

  14. 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

  15. 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)

  16. 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

  17. EPRI BWR Water Chemistry Guidelines Revision

    International Nuclear Information System (INIS)

    Garcia, Susan E.; Giannelli, Joseph F.

    2014-01-01

    BWRVIP-190: BWR Water Chemistry Guidelines – 2008 Revision has been revised. The revision committee consisted of U.S. and non-U.S. utilities (members of the BWR Vessel and Internals Protection (BWRVIP) Mitigation Committee), reactor system manufacturers, fuel suppliers, and EPRI and industry experts. The revised document, BWRVIP-190 Revision 1, was completely reformatted into two volumes, with a simplified presentation of water chemistry control, diagnostic and good practice parameters in Volume 1 and the technical bases in Volume 2, to facilitate use. The revision was developed in parallel and in coordination with preparation of the Fuel Reliability Guidelines Revision 1: BWR Fuel Cladding Crud and Corrosion. Guidance is included for plants operating under normal water chemistry (NWC), moderate hydrogen water chemistry (HWC-M), and noble metal application (GE-Hitachi NobleChem™) plus hydrogen injection. Volume 1 includes significant changes to BWR feedwater and reactor water chemistry control parameters to provide increased assurance of intergranular stress corrosion cracking (IGSCC) mitigation of reactor materials and fuel reliability during all plant conditions, including cold shutdown (≤200°F (93°C)), startup/hot standby (>200°F (93°C) and ≤ 10%) and power operation (>10% power). Action Level values for chloride and sulfate have been tightened to minimize environmentally assisted cracking (EAC) of all wetted surfaces, including those not protected by hydrogen injection, with or without noble metals. Chemistry control guidance has been enhanced to minimize shutdown radiation fields by clarifying targets for depleted zinc oxide (DZO) injection while meeting requirements for fuel reliability. Improved tabular presentations of parameter values explicitly indicate levels at which actions are to be taken and required sampling frequencies. Volume 2 provides the technical bases for BWR water chemistry control for control of EAC, flow accelerated corrosion

  18. Characterization of welding of AISI 304l stainless steel similar to the core encircling of a BWR reactor; Caracterizacion de soldaduras de acero inoxidable AISI 304L similares a las de la envolvente del nucleo de un reactor BWR

    Energy Technology Data Exchange (ETDEWEB)

    Gachuz M, M.E.; Palacios P, F.; Robles P, E.F. [ININ, 52045 Ocoyoacac, Estado de Mexico (Mexico)

    2003-07-01

    Plates of austenitic stainless steel AISI 304l of 0.0381 m thickness were welded by means of the SMAW process according to that recommended in the Section 9 of the ASME Code, so that it was reproduced the welding process used to assemble the encircling of the core of a BWR/5 reactor similar to that of the Laguna Verde Nucleo electric plant, there being generated the necessary documentation for the qualification of the one welding procedure and of the welder. They were characterized so much the one base metal, as the welding cord by means of metallographic techniques, scanning electron microscopy, X-ray diffraction, mechanical essays and fracture mechanics. From the obtained results it highlights the presence of an area affected by the heat of up to 1.5 mm of wide and a value of fracture tenacity (J{sub IC}) to ambient temperature for the base metal of 528 KJ/m{sup 2}, which is diminished by the presence of the welding and by the increment in the temperature of the one essay. Also it was carried out an fractographic analysis of the fracture zone generated by the tenacity essays, what evidence a ductile fracture. The experimental values of resistance and tenacity are important for the study of the structural integrity of the encircling one of the core. (Author)

  19. 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

  20. Hydraulic nuts (HydraNuts) for reactor vessel tensioning

    International Nuclear Information System (INIS)

    Greenwell, Steve

    2008-01-01

    The paper will present how the introduction of hydraulic nuts - HydraNuts, has reduced critical path times, dose exposure for workers and improved working safety conditions around the reactor vessel during tensioning or de-tensioning operations. It will focus upon detailing the advantages realized by utilities that have introduced the technology and providing examples of the improvements made to the process as well as discussing the engineering design change packages required to make the conversion to the new system. HydraNuts replace the traditional mechanical nut/stud tensioning equipment, combining the two functions into a single system, designed for easy installation and operation by one individual. The primary components of the HydraNut can be assembled without the need for external crane or hoist support and are designed so that each sub assembly can be fitted separately. Once all HydraNuts are fitted to the Rx vessel studs and are sitting on the main Rx vessel head flange, then a system of flexible hydraulic hoses is connected to them, forming a closed loop hydraulic harness, which will allow for simultaneous pressurization of all HydraNuts. Hydraulic pressure is obtained by the use of a hydraulic pumping unit and the resultant load generated in each HydraNut is transferred to the stud and main flange closure is obtained. While maintaining hydraulic pressure, a locking ring is rotated into place on the HydraNut assembly that will support the tensioned load mechanically when the hydraulic pressure is released from the hose harness assembly. The hose harness is removed and the HydraNut is now functioning as a mechanical nut retaining the tensioned load. The HydraNut system for Rx vessel applications was first introduced into a plant in the U.S. in October 2006 and based upon the benefits realized subsequent projects are under way within the Asian and U.S. operating fleet. (author)