WorldWideScience

Sample records for hydride reactor vessels

  1. Investigation of Cracked Lithium Hydride Reactor Vessels

    Energy Technology Data Exchange (ETDEWEB)

    bird, e.l.; mustaleski, t.m.

    1999-06-01

    Visual examination of lithium hydride reactor vessels revealed cracks that were adjacent to welds, most of which were circumferentially located in the bottom portion of the vessels. Sections were cut from the vessels containing these cracks and examined by use of the metallograph, scanning electron microscope, and microprobe to determine the cause of cracking. Most of the cracks originated on the outer surface just outside the weld fusion line in the base material and propagated along grain boundaries. Crack depths of those examined sections ranged from {approximately}300 to 500 {micro}m. Other cracks were reported to have reached a maximum depth of 1/8 in. The primary cause of cracking was the creation of high tensile stresses associated with the differences in the coefficients of thermal expansion between the filler metal and the base metal during operation of the vessel in a thermally cyclic environment. This failure mechanism could be described as creep-type fatigue, whereby crack propagation may have been aided by the presence of brittle chromium carbides along the grain boundaries, which indicates a slightly sensitized microstructure.

  2. Reactor vessel

    OpenAIRE

    Makkee, M.; Kapteijn, F.; Moulijn, J.A

    1999-01-01

    A reactor vessel (1) comprises a reactor body (2) through which channels (3) are provided whose surface comprises longitudinal inwardly directed parts (4) and is provided with a catalyst (6), as well as buffer bodies (8, 12) connected to the channels (3) on both sides of the reactor body (2) and comprising connections for supplying (9, 10, 11) and discharging (13, 14, 15) via the channels (3) gases and/or liquids entering into a reaction with each other and substances formed upon this reactio...

  3. Reactor vessel support system. [LMFBR

    Science.gov (United States)

    Golden, M.P.; Holley, J.C.

    1980-05-09

    A reactor vessel support system includes a support ring at the reactor top supported through a box ring on a ledge of the reactor containment. The box ring includes an annular space in the center of its cross-section to reduce heat flow and is keyed to the support ledge to transmit seismic forces from the reactor vessel to the containment structure. A coolant channel is provided at the outside circumference of the support ring to supply coolant gas through the keyways to channels between the reactor vessel and support ledge into the containment space.

  4. 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.

  5. Reactor Structural Materials: Reactor Pressure Vessel Steels

    Energy Technology Data Exchange (ETDEWEB)

    Chaouadi, R

    2000-07-01

    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.

  6. Reactor vessel lower head integrity

    Energy Technology Data Exchange (ETDEWEB)

    Rubin, A.M.

    1997-02-01

    On March 28, 1979, the Three Mile Island Unit 2 (TMI-2) nuclear power plant underwent a prolonged small break loss-of-coolant accident that resulted in severe damage to the reactor core. Post-accident examinations of the TMI-2 reactor core and lower plenum found that approximately 19,000 kg (19 metric tons) of molten material had relocated onto the lower head of the reactor vessel. Results of the OECD TMI-2 Vessel Investigation Project concluded that a localized hot spot of approximately 1 meter diameter had existed on the lower head. The maximum temperature on the inner surface of the reactor pressure vessel (RPV) in this region reached 1100{degrees}C and remained at that temperature for approximately 30 minutes before cooling occurred. Even under the combined loads of high temperature and high primary system pressure, the TMI-2 RPV did not fail. (i.e. The pressure varied from about 8.5 to 15 MPa during the four-hour period following the relocation of melt to the lower plenum.) Analyses of RPV failure under these conditions, using state-of-the-art computer codes, predicted that the RPV should have failed via local or global creep rupture. However, the vessel did not fail; and it has been hypothesized that rapid cooling of the debris and the vessel wall by water that was present in the lower plenum played an important role in maintaining RPV integrity during the accident. Although the exact mechanism(s) of how such cooling occurs is not known, it has been speculated that cooling in a small gap between the RPV wall and the crust, and/or in cracks within the debris itself, could result in sufficient cooling to maintain RPV integrity. Experimental data are needed to provide the basis to better understand these phenomena and improve models of RPV failure in severe accident codes.

  7. Zirconium hydride formation in Hanford production reactor process tubes

    Energy Technology Data Exchange (ETDEWEB)

    Winegardner, W.K.; Griggs, B.

    1967-12-01

    Examination of Zircaloy-2 process tubes from Hanford Production Reactors has revealed extensive zirconium hydride formation. In general, attack is limited to the downstream portions of tubes where aluminum spacers are located. Most of the hydride platelets are contained in a case or layer on the inner surface of the tube. It is not unusual to find cases 0.004 to 0.005 in. thick. Analyses of the 0.037 in. wall tubes with such cases intact often reveal hydrogen concentrations greater than 1000 ppM. Investigation indicates that the hydriding is the result of galvanic contact between aluminum and Zircaloy-2. The galvanic couple (contact between dissimilar metals in the presence of reactor cooling water which serves as the electrolyte) results in the cathodic charging of hydrogen into the Zircaloy.

  8. Optimization of Internal Cooling Fins for Metal Hydride Reactors

    Directory of Open Access Journals (Sweden)

    Vamsi Krishna Kukkapalli

    2016-06-01

    Full Text Available Metal hydride alloys are considered as a promising alternative to conventional hydrogen storage cylinders and mechanical hydrogen compressors. Compared to storing in a classic gas tank, metal hydride alloys can store hydrogen at nearly room pressure and use less volume to store the same amount of hydrogen. However, this hydrogen storage method necessitates an effective way to reject the heat released from the exothermic hydriding reaction. In this paper, a finned conductive insert is adopted to improve the heat transfer in the cylindrical reactor. The fins collect the heat that is volumetrically generated in LaNi5 metal hydride alloys and deliver it to the channel located in the center, through which a refrigerant flows. A multiple-physics modeling is performed to analyze the transient heat and mass transfer during the hydrogen absorption process. Fin design is made to identify the optimum shape of the finned insert for the best heat rejection. For the shape optimization, use of a predefined transient heat generation function is proposed. Simulations show that there exists an optimal length for the fin geometry.

  9. Evaluation of hydride blisters in zirconium pressure tube in CANDU reactor

    Energy Technology Data Exchange (ETDEWEB)

    Cheong, Y. M.; Kim, Y. S.; Gong, U. S.; Kwon, S. C.; Kim, S. S.; Choo, K.N

    2000-09-01

    When the garter springs for maintaining the gap between the pressure tube and the calandria tube are displaced in the CANDU reactor, the sagging of pressure tube results in a contact to the calandria tube. This causes a temperature difference between the inner and outer surface of the pressure tube. The hydride can be formed at the cold spot of outer surface and the volume expansion by hydride dormation causes the blistering in the zirconium alloys. An incident of pressure tube rupture due to the hydride blisters had happened in the Canadian CANDU reactor. This report describes the theoretical development and models on the formation and growth of hydride blister and some experimental results. The evaluation methodology and non-destructive testing for hydride blister in operating reactors are also described.

  10. Shielding efficiency of metal hydrides and borohydrides in fusion reactors

    Directory of Open Access Journals (Sweden)

    Singh Vishvanath P.

    2016-01-01

    Full Text Available Mass attenuation coefficients, mean free paths and exposure buildup factors have been used to characterize the shielding efficiency of metal hydrides and borohydrides, with high density of hydrogen. Gamma ray exposure buildup factors were computed using five-parameter geometric progression fitting at energies 0.015 MeV to15 MeV, and for penetration depths up to 40 mean free paths. Fast-neutron shielding efficiency has been characterized by the effective neutron removal cross-section. It is shown that ZrH2 and VH2 are very good shielding materials for gamma rays and fast neutrons due to their suitable combination of low- and high-Z elements. The present work should be useful for the selection and design of blankets and shielding, and for dose evaluation for components in fusion reactors.

  11. Heat transfer characteristics of the metal hydride vessel based on the plate-fin type heat exchanger

    Science.gov (United States)

    Oi, Tsutomu; Maki, Kohei; Sakaki, Yoshinori

    Heat transfer characteristics of the metal hydride vessel based on the plate-fin type heat exchanger were investigated. Metal hydride beds were filled with AB 2 type hydrogen-storage alloy's particles, Ti 0.42Zr 0.58Cr 0.78Fe 0.57Ni 0.2Mn 0.39Cu 0.03, with a storage capacity of 0.92 wt.%. Heat transfer model in the metal hydride bed based on the heat transfer mechanism for packed bed proposed by Kunii and co-workers is presented. The time-dependent hydrogen absorption/desorption rate and pressure in the metal hydride vessel calculated by the model were compared with the experimental results. During the hydriding, calculated hydrogen absorption rates agreed with measured ones. Calculated thermal equilibrium hydrogen pressures were slightly lower than the measured hydrogen pressures at the inlet of metal hydride vessel. Taking account of the pressure gradient between the inlet of metal hydride vessel and the metal hydride bed, it is considered that this discrepancy is reasonable. During the dehydriding, there were big differences between the calculated hydrogen desorption rates and measured ones. As calculated hydrogen desorption rates were lower than measured ones, there were big differences between the calculated thermal equilibrium hydrogen pressures and the measured hydrogen pressures at the inlet of metal hydride vessel. It is considered that those differences are due to the differences of the heat transfer characteristics such as thermal conductivity of metal hydride particles and porosity between the assumed and actual ones. It is important to obtain the heat transfer characteristics such as thermal conductivity of metal hydride particles and porosity both during the hydriding and dehydriding to design a metal hydride vessel.

  12. Pressure vessel calculations for VVER-440 reactors.

    Science.gov (United States)

    Hordósy, G; Hegyi, Gy; Keresztúri, A; Maráczy, Cs; Temesvári, E; Vértes, P; Zsolnay, E

    2005-01-01

    For the determination of the fast neutron load of the reactor pressure vessel a mixed calculational procedure was developed. The procedure was applied to the Unit II of Paks NPP, Hungary. The neutron source on the outer surfaces of the reactor was determined by a core design code, and the neutron transport calculations outside the core were performed by the Monte Carlo code MCNP. The reaction rate in the activation detectors at surveillance positions and at the cavity were calculated and compared with measurements. In most cases, fairly good agreement was found.

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

    Energy Technology Data Exchange (ETDEWEB)

    Greenspan, Ehud; Todreas, Neil; Taiwo, Temitope

    2009-03-10

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

  14. (Irradiation embrittlement of reactor pressure vessels)

    Energy Technology Data Exchange (ETDEWEB)

    Corwin, W.R.

    1990-09-24

    The traveler served as a member of the two-man US Nuclear Regulatory Commission sponsored team who visited the Prometey Complex in Leningrad to assess the potential for expanded cooperative research concerning integrity of the primary pressure boundary in commercial light-water reactors. The emphasis was on irradiation embrittlement, structural analysis, and fracture mechanics research for reactor pressure vessels. At the irradiation seminar in Cologne, presentations were made by German, French, Finnish, Russian, and US delegations concerning many aspects of irradiation of pressure vessel steels. The traveler made presentations on mechanisms of irradiation embrittlement and on important aspects of the Heavy-Section Steel Irradiation Program results of irradiated fracture mechanics tests.

  15. Reactor Pressure Vessel (RPV) Acquisition Strategy

    Energy Technology Data Exchange (ETDEWEB)

    Mizia, Ronald Eugene [Idaho National Laboratory

    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

  16. Study on the Use of Hydride Fuel in High-Performance Light Water Reactor Concept

    Directory of Open Access Journals (Sweden)

    Haileyesus Tsige-Tamirat

    2015-01-01

    Full Text Available Hydride fuels have features which could make their use attractive in future advanced power reactors. The potential benefit of use of hydride fuel in HPLWR without introducing significant modification in the current core design concept of the high-performance light water reactor (HPLWR has been evaluated. Neutronics and thermal hydraulic analyses were performed for a single assembly model of HPLWR with oxide and hydride fuels. The hydride assembly shows higher moderation with softer neutron spectrum and slightly more uniform axial power distribution. It achieves a cycle length of 18 months with sufficient excess reactivity. At Beginning of Cycle the fuel temperature coefficient of the hydride assembly is higher whereas the moderator and void coefficients are lower. The thermal hydraulic results show that the achievable fuel temperature in the hydride assembly is well below the design limits. The potential benefits of the use of hydride fuel in the current design of the HPLWR with the achieved improvements in the core neutronics characteristics are not sufficient to justify the replacement of the oxide fuel. Therefore for a final evaluation of the use of hydride fuels in HPLWR concepts additional studies which include modification of subassembly and core layout designs are required.

  17. Radiation effects on reactor pressure vessel supports

    Energy Technology Data Exchange (ETDEWEB)

    Johnson, R.E. [Nuclear Regulatory Commission, Washington, DC (United States). Div. of Engineering Technology; Lipinski, R.E. [Idaho National Engineering Lab., Rockville, MD (United States)

    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.

  18. Structural integrity of nuclear reactor pressure vessels

    Science.gov (United States)

    Knott, John F.

    2013-09-01

    The paper starts from concerns expressed by Sir Alan Cottrell, in the early 1970s, related to the safety of the pressurized water reactor (PWR) proposed at that time for the next phase of electrical power generation. It proceeds to describe the design and operation of nuclear generation plant and gives details of the manufacture of PWR reactor pressure vessels (RPVs). Attention is paid to stress-relief cracking and under-clad cracking, experienced with early RPVs, explaining the mechanisms for these forms of cracking and the means taken to avoid them. Particular note is made of the contribution of non-destructive inspection to structural integrity. Factors affecting brittle fracture in RPV steels are described: in particular, effects of neutron irradiation. The use of fracture mechanics to assess defect tolerance is explained, together with the failure assessment diagram embodied in the R6 procedure. There is discussion of the Master Curve and how it incorporates effects of irradiation on fracture toughness. Dangers associated with extrapolation of data to low probabilities are illustrated. The treatment of fatigue-crack growth is described, in the context of transients that may be experienced in the operation of PWR plant. Detailed attention is paid to the thermal shock associated with a large loss-of-coolant accident. The final section reviews the arguments advanced to justify 'Incredibility of Failure' and how these are incorporated in assessments of the integrity of existing plant and proposed 'new build' PWR pressure vessels.

  19. Midland reactor pressure vessel flaw distribution

    Energy Technology Data Exchange (ETDEWEB)

    Foulds, J.R.; Kennedy, E.L. [Failure Analysis Associates, Inc., Menlo Park, CA (United States); Rosinski, S.T. [Sandia National Labs., Albuquerque, NM (United States)

    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.

  20. Reactor pressure vessel structural integrity research

    Energy Technology Data Exchange (ETDEWEB)

    Pennell, W.E.; Corwin, W.R. [Oak Ridge National Lab., TN (United States)

    1995-04-01

    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 shallows 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{sub 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 a 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.

  1. Hygro-Thermo-Mechanical Analysis of a Reactor Vessel

    Directory of Open Access Journals (Sweden)

    Jaroslav Kruis

    2012-01-01

    Full Text Available Determining the durability of a reactor vessel requires a hygro-thermo-mechanical analysis of the vessel throughout its service life. Damage, prestress losses, distribution of heat and moisture and some other quantities are needed for a durability assessment. A coupled analysis was performed on a two-level model because of the huge demands on computer hardware. This paper deals with a hygro-thermo-mechanical analysis of a reactor vessel made of prestressed concrete with a steel inner liner. The reactor vessel is located in Temelín, Czech Republic.

  2. Application of hafnium hydride control rod to large sodium cooled fast breeder reactor

    Energy Technology Data Exchange (ETDEWEB)

    Ikeda, Kazumi, E-mail: kazumi_ikeda@mfbr.mhi.co.jp [Mitsubishi FBR Systems, Inc., 34-17, Jingumae 2-Chome, Shibuya-ku, Tokyo 150-0001 (Japan); Moriwaki, Hiroyuki, E-mail: hiroyuki_moriwaki@mfbr.mhi.co.jp [Mitsubishi FBR Systems, Inc., 34-17, Jingumae 2-Chome, Shibuya-ku, Tokyo 150-0001 (Japan); Ohkubo, Yoshiyuki, E-mail: yoshiyuki_okubo@mfbr.mhi.co.jp [Mitsubishi FBR Systems, Inc., 34-17, Jingumae 2-Chome, Shibuya-ku, Tokyo 150-0001 (Japan); Iwasaki, Tomohiko, E-mail: tomohiko.iwasaki@qse.tohoku.ac.jp [Department of Quantum Science and Energy Engineering, Tohoku University, Aoba, Aramaki, Aoba-ku, Sendai-shi, Miyagi-ken 980-8579 (Japan); Konashi, Kenji, E-mail: konashi@imr.tohoku.ac.jp [Institute for Materials Research, Tohoku University, Narita-cho, Oarai-machi, Higashi-Ibaraki-gun, Ibaraki-ken 311-1313 (Japan)

    2014-10-15

    Highlights: • Application of hafnium hydride control rod to large sodium cooled fast breeder reactor. • This paper treats application of an innovative hafnium hydride control rod to a large sodium cooled fast breeder reactor. • Hydrogen absorption triples the reactivity worth by neutron spectrum shift at H/Hf ratio of 1.3. • Lifetime of the control rod quadruples because produced daughters of hafnium isotopes are absorbers. • Nuclear and thermal hydraulic characteristics of the reactor are as good as or better than B-10 enriched boron carbide. - Abstract: This study treats the feasibility of long-lived hafnium hydride control rod in a large sodium-cooled fast breeder reactor by nuclear and thermal analyses. According to the nuclear calculations, it is found that hydrogen absorption of hafnium triples the reactivity by the neutron spectrum shift at the H/Hf ratio of 1.3, and a hafnium transmutation mechanism that produced daughters are absorbers quadruples the lifetime due to a low incineration rate of absorbing nuclides under irradiation. That is to say, the control rod can function well for a long time because an irradiation of 2400 EFPD reduces the reactivity by only 4%. The calculation also reveals that the hafnium hydride control rod can apply to the reactor in that nuclear and thermal characteristics become as good as or better than 80% B-10 enriched boron carbide. For example, the maximum linear heat rate becomes 3% lower. Owing to the better power distribution, the required flow rate decreases approximately by 1%. Consequently, it is concluded on desk analyses that the long lived hafnium hydride control rod is feasible in the large sodium-cooled fast breeder reactor.

  3. Effect of design parameters on enhancement of hydrogen charging in metal hydride reactors

    Energy Technology Data Exchange (ETDEWEB)

    Kaplan, Y. [Mechanical Engineering Department, Nigde University, 51100 Nigde (Turkey)

    2009-03-15

    The effects of heat transfer mechanisms on the charging process in metal hydride reactors are studied under various charging pressures. Three different cylindrical reactors with the same base dimensions are designed and manufactured. The first one is a closed cylinder cooled with natural convection, the fins are manufactured around the second reactor and the third reactor is cooled with water circulating around the reactor. The temperatures of the reactor at several locations are measured during charging with a range of pressure of 1-10 bar. The third reactor shows the lowest temperature increase with the fastest charging time under all charging pressures investigated. The effective heat transfer coefficients of the reactors are also calculated according to the experimental results and they are found to be 5.5 {+-} 1 W m{sup -2} K{sup -1}, 35 {+-} 2 W m{sup -2} K{sup -1} and 113 {+-} 1 W m{sup -2} K{sup -1}, respectively. The experimental results showed that the charging of hydride reactors is mainly heat transfer dependent and the reactor with better cooling exhibits the fastest charging characteristics. (author)

  4. 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.

  5. Shielding efficiency of metal hydrides and borohydrides in fusion reactors

    OpenAIRE

    Singh Vishvanath P.; Badiger Nagappa M.; Gerward Leif

    2016-01-01

    Mass attenuation coefficients, mean free paths and exposure buildup factors have been used to characterize the shielding efficiency of metal hydrides and borohydrides, with high density of hydrogen. Gamma ray exposure buildup factors were computed using five-parameter geometric progression fitting at energies 0.015 MeV to15 MeV, and for penetration depths up to 40 mean free paths. Fast-neutron shielding efficiency has been characterized by the effective neu...

  6. Measured and calculated fast neutron spectra in a depleted uranium and lithium hydride shielded reactor

    Science.gov (United States)

    Lahti, G. P.; Mueller, R. A.

    1973-01-01

    Measurements of MeV neutron were made at the surface of a lithium hydride and depleted uranium shielded reactor. Four shield configurations were considered: these were assembled progressively with cylindrical shells of 5-centimeter-thick depleted uranium, 13-centimeter-thick lithium hydride, 5-centimeter-thick depleted uranium, 13-centimeter-thick lithium hydride, 5-centimeter-thick depleted uranium, and 3-centimeter-thick depleted uranium. Measurements were made with a NE-218 scintillation spectrometer; proton pulse height distributions were differentiated to obtain neutron spectra. Calculations were made using the two-dimensional discrete ordinates code DOT and ENDF/B (version 3) cross sections. Good agreement between measured and calculated spectral shape was observed. Absolute measured and calculated fluxes were within 50 percent of one another; observed discrepancies in absolute flux may be due to cross section errors.

  7. Design for the WWR-M reactor vessel removal

    Energy Technology Data Exchange (ETDEWEB)

    Lobach, Yu.N., E-mail: lobach@kinr.kiev.ua [Institute for Nuclear Research of NASU, Prospekt Nauki 47, 03680 Kiev (Ukraine); Toth, G., E-mail: gtoth@aeki.kfki.hu [Centre for Energy Research of HAS, Konkoly Thege Miklós út 29-33, Budapest 1121 (Hungary)

    2013-05-15

    Highlights: ► The current status of the decommissioning planning for the WWR-M reactor is outlined. ► The general dismantling strategy consists of the dismantling and removal of the separate bulky elements as whole pieces without preliminary segmentation. ► The design for the reactor vessel extraction was selected and developed. -- Abstract: The final decommissioning planning for the Kiev's research reactor WWR-M is in progress now. The general dismantling strategy consists of the dismantling and removal of the separate bulky elements as whole pieces without preliminary segmentation. The reactor vessel removal is considered as the key element in the sequence of dismantling works; a separate design for the reactor vessel extraction was developed. The extensive analysis focused on the optimization of this technical task has been performed. The outline of the design is presented in this paper.

  8. EXPERIMENTAL RESULTS FOR THE ISOTOPIC EXCHANGE OF A 1600 LITER TITANIUM HYDRIDE STORAGE VESSEL

    Energy Technology Data Exchange (ETDEWEB)

    Klein, J.

    2010-12-14

    Titanium is used as a low pressure tritium storage material. The absorption/desorption rates and temperature rise during air passivation have been reported previously for a 4400 gram prototype titanium hydride storage vessel (HSV). A desorption limit of roughly 0.25 Q/M was obtained when heating to 700 C which represents a significant residual tritium process vessel inventory. To prepare an HSV for disposal, batchwise isotopic exchange has been proposed to reduce the tritium content to acceptable levels. A prototype HSV was loaded with deuterium and exchanged with protium to determine the effectiveness of a batch-wise isotopic exchange process. A total of seven exchange cycles were performed. Gas samples were taken nominally at the beginning, middle, and end of each desorption cycle. Sample analyses showed the isotopic exchange process does not follow the standard dilution model commonly reported. Samples taken at the start of the desorption process were lower in deuterium (the gas to be removed) than those taken later in the desorption cycle. The results are explained in terms of incomplete mixing of the exchange gas in the low pressure hydride.

  9. Design and analysis of multicavity prestressed concrete reactor vessels. [HTGR

    Energy Technology Data Exchange (ETDEWEB)

    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.

  10. The survey of Neutron moderating properties of zirconium hydride nanoparticles (ZrH2 in the reactors of nuclear powerhouse

    Directory of Open Access Journals (Sweden)

    Ahmad Nozad Golikand

    2017-01-01

    Full Text Available Metal hydrides as a Neutron Moderator (NMs have effective and impressive application in nuclear reactors. Unquestionably, Retarder should be close to the atomic mass of the neutron to be able to reduce its energy with no interaction with the neutrons. The hydrogen atom nucleons have the atomic Mass close to the Neutron. Surprisingly, Metal hydrides can absorb a high percentage of hydrogen. Metal Hydrides have very good properties at high temperatures and can also maintain it even at higher temperatures. Due to the high volume of hydrogen stored in the metal hydrides, they blurt their good mechanical properties and Moderating Effect out on the incidence of the various reactors. Since they have ultrahigh functionality of Hydrogen storage, and likewise, they reveal their unique neutron moderating and thermal properties, in the present research, it was scrutinized the feasibility and circumstance of using them in the nuclear reactors of atomic energy powerhouse. Hence, in order to sonochemical reaction synthesis of zirconium hydride noparticles (ZrH2, the precursor of zirconium tetrachloride (ZrCl4 and potassium hydroxide (KOH were utilized as neutralizing acidic environment, exposure to the ultrasound waves. Eventually, some diverse metal salts were come on competitive comparison with zirconium hydride salt that the zirconium hydride nanoparticles had the best performance as compared to all of them.

  11. Reactor pressure vessels as type B transport containment boundaries

    Energy Technology Data Exchange (ETDEWEB)

    Nickell, R.E. [Applied Science and Technology, Inc., Poway, CA (United States); Griesbach, T.J. [ATI Consulting, Danville, CA (United States)

    1998-07-01

    Transportation risk and personnel exposure, as well as the cost of decommissioning nuclear power plants, can all be reduced significantly through the one-time use of the reactor pressure vessel as a containment boundary for shipping the activated internal components from the reactor site to a burial site. In order to help provide the technical basis for this end-use application, the ASME Board on Nuclear Codes and Standards, through its Subcommittee XI, has prepared a draft nuclear code case that contains requirements for any modifications to the vessel, including materials, design, fabrication, and examination. In particular, the requirements for evaluation of potential brittle fracture as the result of potentially low ambient shipping temperatures combined with hypothetical transportation accident loading are addressed. Existing ASME Code Section XI rules for linear elastic fracture mechanics evaluation of irradiated reactor pressure vessels have been adapted and included in the code case. (authors)

  12. ANALISA PENGARUH PERIODIK CHARGING DAN DISCHARGING PADA WAKTU CHARGING DAN DISCHARGING DARI METAL HYDRIDE REACTOR (MHR

    Directory of Open Access Journals (Sweden)

    Taurista Perdana Syawitri

    2015-01-01

    Full Text Available Perpindahan panas merupakan faktor penting yang mempengaruhi kinerja tangki penyimpanan hydrogen dalam bentuk metal hydride. Penelitian sebelumnya menyimpulkan bahwa penambahan sejumlah kecil metal foam pada interior metal hydride reactor (MHR merupakan cara yang efektif untuk meningkatkan perpindahan panas pada reaktor, sehingga mengurangi waktu charging. Namun, proses charging dan discharging MHR dilakukan secara terpisah dalam studi ini. Oleh karena itu, penelitian ini menyimulasikan model 2-D axisymmetric menggunakan software COMSOL untuk menginvestigasi pengaruh dari periodik charging dan discharging dari MHR yang mengandung metal foam dengan fraksi volume tertentu. Hasil simulasi menunjukkan bahwa model saat ini mempunyai hasil yang sesuai dengan hasil yang ditunjukkan di dalam literatur. Selain itu, ditemukan bahwa terdapat nilai optimum dari fraksi volume metal foam di setiap proses charging dan proses discharging yang memberikan waktu charging dan discharging lebih pendek.

  13. 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.

  14. PWR reactor vessel in-service-inspection according to RSEM

    Energy Technology Data Exchange (ETDEWEB)

    Algarotti, Marc; Dubois, Philippe; Hernandez, Luc; Landez, Jean Paul [Intercontrole, 13, rue du Capricorne - SILIC 433, 94583 Rungis - Cedex (France)

    2006-07-01

    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

  15. Ex-vessel Steam Explosion Analysis for Pressurized Water Reactor and Boiling Water Reactor

    OpenAIRE

    Matjaž Leskovar; Mitja Uršič

    2016-01-01

    A steam explosion may occur during a severe accident, when the molten core comes into contact with water. The pressurized water reactor and boiling water reactor ex-vessel steam explosion study, which was carried out with the multicomponent three-dimensional Eulerian fuel–coolant interaction code under the conditions of the Organisation for Economic Co-operation and Development (OECD) Steam Explosion Resolution for Nuclear Applications project reactor exercise, is presented and discussed. In ...

  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. Retrospective dosimetry analyses of reactor vessel cladding samples

    Energy Technology Data Exchange (ETDEWEB)

    Greenwood, L. R.; Soderquist, C. Z. [Battelle Pacific Northwest National Laboratory, Richland, WA 99352 (United States); Fero, A. H. [Westinghouse Electric Company, Cranberry Twp., PA 16066 (United States)

    2011-07-01

    Reactor pressure vessel cladding samples for Ringhals Units 3 and 4 in Sweden were analyzed using retrospective reactor dosimetry techniques. The objective was to provide the best estimates of the neutron fluence for comparison with neutron transport calculations. A total of 51 stainless steel samples consisting of chips weighing approximately 100 to 200 mg were removed from selected locations around the pressure vessel and were sent to Pacific Northwest National Laboratory for analysis. The samples were fully characterized and analyzed for radioactive isotopes, with special interest in the presence of Nb-93m. The RPV cladding retrospective dosimetry results will be combined with a re-evaluation of the surveillance capsule dosimetry and with ex-vessel neutron dosimetry results to form a comprehensive 3D comparison of measurements to calculations performed with 3D deterministic transport code. (authors)

  18. Numerical simulation of hydrogen desorption from high-density metal hydride hydrogen storage vessels

    Science.gov (United States)

    Sang-Kun, O.; Yi, Kyung-Woo; Cho, Sung-Wook

    2017-07-01

    Metal hydride (MH) alloys are a promising type of material in hydrogen storage applications, allowing for low-pressure, high-density storage. However, while many studies are being performed on enhancing the hydrogen storage properties of such alloys, there has been little research on large-scale storage vessels which make use of the alloys. In particular, large-scale, high-density storage devices must make allowances for the temperature variations caused by the heat of reaction between hydrogen and MH alloys, which may impact the storage characteristics. In this study, we propose a numerical model for the design and evaluation of hydrogen storage devices using MH alloys. Hydrogen desorption reaction behavior for an alloy is observed in terms of temperature and reaction rate. This behavioral correlation is used as the basis for a comprehensive simulation model of the alloy system. Calculated results are found to be in good agreement with experimentally measured data, indicating that the model may be applied to multiple system geometries, scales, and alloy compositions.

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

    Energy Technology Data Exchange (ETDEWEB)

    Cui, J.; Shek, G.K. [Kinectrics Inc., Toronto, Ontario (Canada); Wang, Z.R. [Toronto Univ., Dept. of Materials Science and Engineering, Toronto, Ontario (Canada)

    2007-07-01

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

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

    Science.gov (United States)

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

    2008-12-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

    Vaibhaw, Kumar [Nuclear Fuel Complex, ECIL Post, Hyderabad 500 062 (India)], E-mail: krvaibhaw@yahoo.co.in; Rao, S.V.R.; Jha, S.K.; Saibaba, N.; Jayaraj, R.N. [Nuclear Fuel Complex, ECIL Post, Hyderabad 500 062 (India)

    2008-12-15

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

  2. 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.

  3. EPRI activities to address reactor pressure vessel integrity issues

    Energy Technology Data Exchange (ETDEWEB)

    Rosinski, S.T.; Carter, R.G. [Electric Power Res. Inst., Charlotte, NC (United States)

    1999-12-01

    The demonstration of reactor pressure vessel (RPV) structural integrity is an essential element in ensuring the continued safe and reliable operation of US nuclear power plants. The Electric Power Research Institute (EPRI), through its domestic and international member utilities, continues to pursue an aggressive research program to develop technologies and capabilities that will address issues associated with reactor pressure vessel integrity. Ongoing research in the EPRI nuclear power group materials performance program covers a broad range of technical areas associated with RPVs. The program is structured under the following product groups; (1) management and mitigation; (2) material performance databases; (3) material condition assessment; and (4) operability assessment. Specific activities under each of theses product groups are described in this paper. (orig.)

  4. Annealing the reactor vessel at the Palisades Plant

    Energy Technology Data Exchange (ETDEWEB)

    Fenech, R.A. [Palisades Plant, Covert, MI (United States)

    1996-03-01

    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.

  5. Fatigue Test of Domestic Manufactured Reactor Pressure Vessel Steel

    Institute of Scientific and Technical Information of China (English)

    ZHONG; Wei-hua; TONG; Zhen-feng; NING; Guang-sheng; YU; Bin-tao

    2013-01-01

    The CAP1400 will be built by our country,after the self-dependent innovation work on the imported technology of AP1000,which is a 3rd generation NPP.Now,the design of CAP1400 key equipment is ongoing,and the fatigue design of the domestic manufactured key equipment,such as reactor pressure vessel(RPV),is found to be a main problem in the design work,as the fatigue data is lacked.Thus the

  6. Thermally activated deformation of irradiated reactor pressure vessel steel

    Science.gov (United States)

    Böhmert, J.; Müller, G.

    2002-03-01

    Temperature and strain rate change tensile tests were performed on two VVER 1000-type reactor pressure vessel welds with different contents of nickel in unirradiated and irradiated conditions in order to determine the activation parameters of the contribution of the thermally activated deformation. There are no differences of the activation parameters in the unirradiated and the irradiated conditions as well as for the two different materials. This shows that irradiation hardening preferentially results from a friction hardening mechanism by long-range obstacles.

  7. Development of ROV System for FOSAR in Reactor Vessel

    Energy Technology Data Exchange (ETDEWEB)

    Choi, Young Soo; Kim, Tae Won; Lee, Sung Uk; Jeong, Kyung Min [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Kim, Nam Kyun [Korea Plant Service and Engineering Co., Seongnam (Korea, Republic of)

    2011-10-15

    Foreign object in the reactor vessel is susceptible to damage the fuel. Prior to reloading fuel assemblies into the core, FOSAR(Foreign Object Search And Retrieval) activities were performed on and beneath the lower core plate with conventional equipment. However, the reactor vessel is limited to humans who are susceptible to radiation exposure, and conventional equipment is hard to access because of the complexity of the structure. To improve the convenience of use and retrieval ability in the under-core plate region, we are developing a FOSAR system carried by ROV (Remotely Operated Vehicle). In this paper, we describe a ROV system developed. The ROV system is composed of robot vehicle and remote control unit. The vehicle has 4 thrusters, tilt, camera, light and depth sensor, etc. Considering radiation damage, processors are not equipped on the vehicle. Control signals and sensing signals are transferred through umbilical cable. Remote control unit is composed of electric driving module and two computers which one is for the control and the other is for the detection of robot position. Control computer has a joystick user input and video/signal input, and transmit motor control signal and lens control signal via CAN/RS485 communication. And the other computers transmit information of vehicle position to the control computer via serial communication. Information of vehicle position is obtained through image processing algorithm. The acquiring camera of vehicle is on the flange of reactor vessel. Simulations on the detection of vehicle position are performed at the reactor vessel mockup which scaled down by 6 and verified to use in the control of robot by visual tracking. And functional test has been performed on the air condition. In the future, performance test will be carried out real sized mockup and underwater condition

  8. Neutron flux reduction programs for reactor pressure vessel

    Energy Technology Data Exchange (ETDEWEB)

    Yoo, C.S. [Korea Atomic Energy Research Inst. KAERI, 150 Deogjin-dong, Yuseong-gu, Daejeon 305-353 (Korea, Republic of); Kim, B.C. [Korea Reactor Integrity Surveillance Technology KRIST, 150 Deogjin-dong, Yuseong-gu, Daejeon 305-353 (Korea, Republic of)

    2011-07-01

    The objective of this work is to implement various fast neutron flux reduction programs on the belt-line region of the reactor pressure vessel to reduce the increasing rate of reference temperature for pressurized thermal shock (RT PTS) for Korea Nuclear Unit 1. A pressurized thermal shock (PTS) event is an event or transient in pressurized water reactors (PWRs) causing severe overcooling (thermal shock) concurrent with or followed by significant pressure in the reactor vessel. A PTS concern arises if one of these transients acts in the belt-line region of a reactor vessel where a reduced fracture resistance exists because of neutron irradiation. Generally, the RT PTS value is continuously increasing according to the fast neutron irradiation during the reactor operation, and it can reach the screening criterion prior to the expiration of the operating license. To reduce the increasing rate of RT PTS, various neutron flux reduction programs can be implemented, which are focused on license renewal. In this paper, neutron flux reduction programs, such as low leakage loading pattern strategy, loading of neutron absorber rods, and dummy fuel assembly loading are considered for Korea Nuclear Unit 1, of which the RT PTS value of the leading material (circumferential weld) is going to reach the screening criterion in the near future. To evaluate the effects of the neutron flux reduction programs, plant and cycle specific forward neutron transport calculations for the various neutron flux reduction programs were carried out. For the analysis, all transport calculations were carried out by using the DORT 3.1 discrete ordinate code and BUGLE-96 cross-section library. (authors)

  9. 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.

  10. REACTOR PRESSURE VESSEL TEMPERATURE ANALYSIS OF CANDIDATE VERY HIGH TEMPERATURE REACTOR DESIGNS

    Energy Technology Data Exchange (ETDEWEB)

    Hans D. Gougar; Cliff B. Davis; George Hayner; Kevan Weaver

    2006-10-01

    Analyses were performed to determine maximum temperatures in the reactor pressure vessel for two potential Very-High Temperature Reactor (VHTR) designs during normal operation and during a depressurized conduction cooldown accident. The purpose of the analyses was to aid in the determination of appropriate reactor vessel materials for the VHTR. The designs evaluated utilized both prismatic and pebble-bed cores that generated 600 MW of thermal power. Calculations were performed for fluid outlet temperatures of 900 and 950 °C, corresponding to the expected range for the VHTR. The analyses were performed using the RELAP5-3D and PEBBED-THERMIX computer codes. Results of the calculations were compared with preliminary temperature limits derived from the ASME pressure vessel code. Because PEBBED-THERMIX has not been extensively validated, confirmatory calculations were also performed with RELAP5-3D for the pebble-bed design. During normal operation, the predicted axial profiles in reactor vessel temperature were similar with both codes and the predicted maximum values were within 2 °C. The trends of the calculated vessel temperatures were similar during the depressurized conduction cooldown accident. The maximum value predicted with RELAP5-3D during the depressurized conduction cooldown accident was about 40 °C higher than that predicted with PEBBED. This agreement is considered reasonable based on the expected uncertainty in either calculation. The differences between the PEBBED and RELAP5-3D calculations were not large enough to affect conclusions concerning comparisons between calculated and allowed maximum temperatures during normal operation and the depressurized conduction cooldown accident.

  11. Ex-vessel Steam Explosion Analysis for Pressurized Water Reactor and Boiling Water Reactor

    Directory of Open Access Journals (Sweden)

    Matjaž Leskovar

    2016-02-01

    Full Text Available A steam explosion may occur during a severe accident, when the molten core comes into contact with water. The pressurized water reactor and boiling water reactor ex-vessel steam explosion study, which was carried out with the multicomponent three-dimensional Eulerian fuel–coolant interaction code under the conditions of the Organisation for Economic Co-operation and Development (OECD Steam Explosion Resolution for Nuclear Applications project reactor exercise, is presented and discussed. In reactor calculations, the largest uncertainties in the prediction of the steam explosion strength are expected to be caused by the large uncertainties related to the jet breakup. To obtain some insight into these uncertainties, premixing simulations were performed with both available jet breakup models, i.e., the global and the local models. The simulations revealed that weaker explosions are predicted by the local model, compared to the global model, due to the predicted smaller melt droplet size, resulting in increased melt solidification and increased void buildup, both reducing the explosion strength. Despite the lower active melt mass predicted for the pressurized water reactor case, pressure loads at the cavity walls are typically higher than that for the boiling water reactor case. This is because of the significantly larger boiling water reactor cavity, where the explosion pressure wave originating from the premixture in the center of the cavity has already been significantly weakened on reaching the distant cavity wall.

  12. Use of MCNP for characterization of reactor vessel internals waste from decommissioned nuclear reactors

    Energy Technology Data Exchange (ETDEWEB)

    Love, E.F.; Pauley, K.A.; Reid, B.D.

    1995-09-01

    This study describes the use of the Monte Carlo Neutron-Photon (MCNP) code for determining activation levels of irradiated reactor vessel internals hardware. The purpose of the analysis is to produce data for the Department of Energy`s Greater-Than-Class C Low-Level Radioactive Waste Program. An MCNP model was developed to analyze the Yankee Rowe reactor facility. The model incorporates reactor geometry, material compositions, and operating history data acquired from Yankee Atomic Electric Company. In addition to the base activation analysis, parametric studies were performed to determine the sensitivity of activation to specific parameters. A component sampling plan was also developed to validate the model results, although the plan was not implemented. The calculations for the Yankee Rowe reactor predict that only the core baffle and the core support plates will be activated to levels above the Class C limits. The parametric calculations show, however, that the large uncertainties in the material compositions could cause errors in the estimates that could also increase the estimated activation level of the core barrel to above the Class C limits. Extrapolation of the results to other reactor facilities indicates that in addition to the baffle and support plates, core barrels may also be activated to above Class C limits; however the classification will depend on the specific operating conditions of the reactor and the specific material compositions of the metal, as well as the use of allowable concentration averaging practices in packaging and classifying the waste.

  13. Reactor vessel water level estimation during severe accidents using cascaded fuzzy neural networks

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Dong Yeong; Yoo, Kwae Hwan; Choi, Geon Pil; Back, Ju Hyun; Na, Man Gyun [Dept. of Nuclear Engineering, Chosun University, Gwangju (Korea, Republic of)

    2016-06-15

    Global concern and interest in the safety of nuclear power plants have increased considerably since the Fukushima accident. In the event of a severe accident, the reactor vessel water level cannot be measured. The reactor vessel water level has a direct impact on confirming the safety of reactor core cooling. However, in the event of a severe accident, it may be possible to estimate the reactor vessel water level by employing other information. The cascaded fuzzy neural network (CFNN) model can be used to estimate the reactor vessel water level through the process of repeatedly adding fuzzy neural networks. The developed CFNN model was found to be sufficiently accurate for estimating the reactor vessel water level when the sensor performance had deteriorated. Therefore, the developed CFNN model can help provide effective information to operators in the event of a severe accident.

  14. Low Temperature Irradiation Embrittlement of Reactor Pressure Vessel Steels

    Energy Technology Data Exchange (ETDEWEB)

    Wang, Jy-An John [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2015-08-01

    The embrittlement trend curve development project for HFIR reactor pressure vessel (RPV) steels was carried out with three major tasks. Which are (1) data collection to match that used in HFIR steel embrittlement trend published in 1994 Journal Nuclear Material by Remec et. al, (2) new embrittlement data of A212B steel that are not included in earlier HFIR RPV trend curve, and (3) the adjustment of nil-ductility-transition temperature (NDTT) shift data with the consideration of the irradiation temperature effect. An updated HFIR RPV steel embrittlement trend curve was developed, as described below. NDTT( C) = 23.85 log(x) + 203.3 log (x) + 434.7, with 2- uncertainty of 34.6 C, where parameter x is referred to total dpa. The developed update HFIR RPV embrittlement trend curve has higher embrittlement rate compared to that of the trend curve developed in 1994.

  15. Subsize specimen testing of nuclear reactor pressure vessel material

    Energy Technology Data Exchange (ETDEWEB)

    Kumar, A.S. [Missouri Univ., Rolla, MO (United States). Materials Research Center; Rosinski, S.T. [Sandia National Labs., Albuquerque, NM (United States); Cannon, N.S. [Westinghouse Hanford Co., Richland, WA (United States); Hamilton, M.L. [Pacific Northwest Lab., Richland, WA (United States)

    1991-12-31

    A new methodology is proposed to correlate the upper shelf energy (USE) of full size and subsize Charpy specimens of a nuclear reactor pressure vessel plate material, A533B. The methodology appears to be more satisfactory than the methodologies proposed earlier. USE of a notched-only specimen is partitioned into macro-crack initiation and crack propagation energies. USE of a notched and precracked specimen provides the crack propagation energy. {Delta}USE, the difference between the USE`s of notched-only and precracked specimens, is an estimate of the crack initiation energy. {Delta}USE was normalized by a factor involving the dimensions of the Charpy specimen and the stress concentration factor at the notch root. The normalized values of the {Delta}USE were found to be invariant with specimen size.

  16. Subsize specimen testing of nuclear reactor pressure vessel material

    Energy Technology Data Exchange (ETDEWEB)

    Kumar, A.S. (Missouri Univ., Rolla, MO (United States). Materials Research Center); Rosinski, S.T. (Sandia National Labs., Albuquerque, NM (United States)); Cannon, N.S. (Westinghouse Hanford Co., Richland, WA (United States)); Hamilton, M.L. (Pacific Northwest Lab., Richland, WA (United States))

    1991-01-01

    A new methodology is proposed to correlate the upper shelf energy (USE) of full size and subsize Charpy specimens of a nuclear reactor pressure vessel plate material, A533B. The methodology appears to be more satisfactory than the methodologies proposed earlier. USE of a notched-only specimen is partitioned into macro-crack initiation and crack propagation energies. USE of a notched and precracked specimen provides the crack propagation energy. [Delta]USE, the difference between the USE's of notched-only and precracked specimens, is an estimate of the crack initiation energy. [Delta]USE was normalized by a factor involving the dimensions of the Charpy specimen and the stress concentration factor at the notch root. The normalized values of the [Delta]USE were found to be invariant with specimen size.

  17. Prestressed concrete reactor vessel thermal cylinder model study

    Energy Technology Data Exchange (ETDEWEB)

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

    1977-05-04

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

  18. 77 FR 23513 - Updated Aging Management Criteria for Reactor Vessel Internal Components of Pressurized Water...

    Science.gov (United States)

    2012-04-19

    ... COMMISSION Updated Aging Management Criteria for Reactor Vessel Internal Components of Pressurized Water... Management Criteria for PWR Reactor Vessel Internal Components.'' The original notice provided the ADAMS... published a notice requesting public comments on draft LR-ISG-2011-04, ``Updated Aging Management...

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2001-03-15

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

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

    Energy Technology Data Exchange (ETDEWEB)

    Huh, Hoon; Chang, Soon Heung; Choi, Tae Hoon; Kim, Hyun Sop; Yang, Soo Hyung; Kim, Soo Hyung; Kim, Seung Hop; An Hyung Taek; Jeong, Yong Hoon; Huh, Gyun Young [Korea Advanced Institute of Science and Technology, Taejon (Korea, Republic of)

    1998-03-15

    Cooling methodologies for the molten corium resulted from the severe accident of the nuclear power plant is suggested as one of most important items for the safety of the NPP. In this regard, considerable experimental and analytical works have been devoted. In the 1st phase of this project, present status related to the external reactor vessel cooling for the retention of the corium in the reactor vessel and corium at the reactor cavity have been investigated and preliminary studies have been accomplished for the detail evaluation of the each cooling methodology. The preliminary studies include the analysis and detail investigation of the possible phenomena, investigation of the heat transfer correlations and preliminary evaluation of the external reactor vessel cooling using the developed computer code.

  1. SAFT inspections for developing empirical database of fabrication flaws in nuclear reactor pressure vessels

    Science.gov (United States)

    Doctor, Steven R.; Schuster, George J.; Pardini, Allan F.

    1998-03-01

    The Pacific Northwest National Laboratory (PNNL) is developing a methodology for estimating the size and density distribution of fabrication flaws in U.S. nuclear reactor pressure vessels. This involves the nondestructive evaluation (NDE) of reactor pressure vessel materials and the destructive validation of the flaws found. NDE has been performed on reactor pressure vessel material made by Babcock & Wilcox and Combustion Engineering. A metallographic analysis is being performed to validate the flaw density and size distributions estimated from the 2500 indications of fabrication flaws that were detected and characterized in the very sensitive SAFT-UT (synthetic aperture focusing technique for ultrasonic testing) inspection data from the Pressure Vessel Research User Facility (PVRUF) vessel at Oak Ridge National Laboratory. Research plans are also described for expanding the work to include other reactor pressure vessel materials.

  2. Embrittlement recovery due to annealing of reactor pressure vessel steels

    Energy Technology Data Exchange (ETDEWEB)

    Eason, E.D.; Wright, J.E.; Nelson, E.E. [Modeling and Computing Services, Boulder, CO (United States); Odette, G.R.; Mader, E.V. [Univ. of California, Santa Barbara, CA (United States)

    1996-03-01

    Embrittlement of reactor pressure vessels (RPVs) can be reduced by thermal annealing at temperatures higher than the normal operating conditions. Although such an annealing process has not been applied to any commercial plants in the United States, one US Army reactor, the BR3 plant in Belgium, and several plants in eastern Europe have been successfully annealed. All available Charpy annealing data were collected and analyzed in this project to develop quantitative models for estimating the recovery in 30 ft-lb (41 J) Charpy transition temperature and Charpy upper shelf energy over a range of potential annealing conditions. Pattern recognition, transformation analysis, residual studies, and the current understanding of the mechanisms involved in the annealing process were used to guide the selection of the most sensitive variables and correlating parameters and to determine the optimal functional forms for fitting the data. The resulting models were fitted by nonlinear least squares. The use of advanced tools, the larger data base now available, and insight from surrogate hardness data produced improved models for quantitative evaluation of the effects of annealing. The quality of models fitted in this project was evaluated by considering both the Charpy annealing data used for fitting and the surrogate hardness data base. The standard errors of the resulting recovery models relative to calibration data are comparable to the uncertainty in unirradiated Charpy data. This work also demonstrates that microhardness recovery is a good surrogate for transition temperature shift recovery and that there is a high level of consistency between the observed annealing trends and fundamental models of embrittlement and recovery processes.

  3. Evaluation on Waste Volume and Weight from Decommissioning of Kori Unit 1 Reactor Vessel

    Energy Technology Data Exchange (ETDEWEB)

    Choi, Yujeong; Lee, Seong-Cheol; Kim, Chang-Lak [KEPCO International Nuclear Graduate School, Ulsan (Korea, Republic of)

    2015-05-15

    In this paper, the concept of cutting reactor vessel and container for decommissioning Kori unit 1 has been investigated. As a result of the investigation, it is found that cutting the reactor vessel into small pieces, especially for upper and bottom heads of the reactor vessel, is more effective to reduce total disposal volume generated from decommissioning. As a part of continuing efforts to prepare shut down of nuclear power plant, several researches have been conducted to establish plans to dispose decommissioning waste from nuclear power plants. When decommissioning nuclear power plant, most of radioactive waste is generated from primary side including a reactor vessel. Radioactive waste amounts generated from decommissioning is significantly affected by several factors, such as dismantling method, waste classification, reactor lifetime, disposal method and etc.

  4. Miniaturized Charpy test for reactor pressure vessel embrittlement characterization

    Energy Technology Data Exchange (ETDEWEB)

    Manahan, M.P. Sr. [MPM Research and Consulting, Lemont, PA (United States)

    1999-10-01

    Modifications were made to a conventional Charpy machine to accommodate the miniaturized Charpy V-Notch (MCVN) specimens which were fabricated from an archived reactor pressure vessel (RPV) steel. Over 100 dynamic MCVN tests were performed and compared to the results from conventional Charpy V-Notch (CVN) tests to demonstrate the efficacy of the miniature specimen test. The optimized sidegrooved MCVN specimens exhibit transitional fracture behavior over essentially the same temperature range as the CVN specimens which indicates that the stress fields in the MCVN specimens reasonably simulate those of the CVN specimens and this fact has been observed in finite element calculations. This result demonstrates a significant breakthrough since it is now possible to measure the ductile-brittle transition temperature (DBTT) using miniature specimens with only small correction factors, and for some materials as in the present study, without the need for any correction factor at all. This development simplifies data interpretation and will facilitate future regulatory acceptance. The non-sidegrooved specimens yield energy-temperature data which is significantly shifted downward in temperature (non-conservative) as a result of the loss of constraint which accompanies size reduction.

  5. Weld repair of helium degraded reactor vessel material

    Energy Technology Data Exchange (ETDEWEB)

    Kanne, W.R. Jr.; Lohmeier, D.A.; Louthan, M.R. Jr.; Rankin, D.T.; Franco-Ferreira, E.A. (Westinghouse Savannah River Co., Aiken, SC (United States)); Bruck, G.J.; Madeyski, A.; Shogan, R.P.; Lessmann, G.G. (Westinghouse Electric Corp., Pittsburgh, PA (United States). Science and Technology Center)

    1990-01-01

    Welding methods for modification or repair of irradiated nuclear reactor vessels are being evaluated at the Savannah River Site. A low-penetration weld overlay technique has been developed to minimize the adverse effects of irradiation induced helium on the weldability of metals and alloys. This technique was successfully applied to Type 304 stainless steel test plates that contained 3 to 220 appm helium from tritium decay. Conventional welding practices caused significant cracking and degradation in the test plates. Optical microscopy of weld surfaces and cross sections showed that large surface toe cracks formed around conventional welds in the test plates but did not form around overlay welds. Scattered incipient underbead cracks (grain boundary separations) were associated with both conventional and overlay test welds. Tensile and bend tests were used to assess the effect of base metal helium content on the mechanical integrity of the low-penetration overlay welds. The axis of tensile specimens was perpendicular to the weld-base metal interface. Tensile specimens were machined after studs were resistance welded to overlay surfaces.

  6. Weld repair of helium degraded reactor vessel material

    Energy Technology Data Exchange (ETDEWEB)

    Kanne, W.R. Jr.; Lohmeier, D.A.; Louthan, M.R. Jr.; Rankin, D.T.; Franco-Ferreira, E.A. [Westinghouse Savannah River Co., Aiken, SC (United States); Bruck, G.J.; Madeyski, A.; Shogan, R.P.; Lessmann, G.G. [Westinghouse Electric Corp., Pittsburgh, PA (United States). Science and Technology Center

    1990-12-31

    Welding methods for modification or repair of irradiated nuclear reactor vessels are being evaluated at the Savannah River Site. A low-penetration weld overlay technique has been developed to minimize the adverse effects of irradiation induced helium on the weldability of metals and alloys. This technique was successfully applied to Type 304 stainless steel test plates that contained 3 to 220 appm helium from tritium decay. Conventional welding practices caused significant cracking and degradation in the test plates. Optical microscopy of weld surfaces and cross sections showed that large surface toe cracks formed around conventional welds in the test plates but did not form around overlay welds. Scattered incipient underbead cracks (grain boundary separations) were associated with both conventional and overlay test welds. Tensile and bend tests were used to assess the effect of base metal helium content on the mechanical integrity of the low-penetration overlay welds. The axis of tensile specimens was perpendicular to the weld-base metal interface. Tensile specimens were machined after studs were resistance welded to overlay surfaces.

  7. Development of automatic Ultrasonic testing equipment for reactor pressure vessel

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Kor R.; Kim, Jae H.; Lee, Jae C.

    1996-06-01

    The selected weld areas of a reactor pressure vessel and adjacent piping are examined by the remote mechanized ultrasonic testing (MUT) equipment. Since the MUT equipment was purchased from southwest Research Institute (SwRI) in April 1985, 15 inservice inspections and 5 preservice inspections are performed with this MUT equipment. However due to the old age of the equipment and frequent movements to plant sites, the reliability of examination was recently decreased rapidly and it is very difficult to keep spare parts. In order to resolve these problems and to meet the strong request from plant sites, we intend to develop a new 3-axis control system including hardware and software. With this control system, we expect more efficient and reliable examination of the nozzle to shell weld areas, which is specified in ASME Code Section XI. The new 3-axis control system hardware and software were designed and development of our own control system, the advanced technologies of computer control mechanism were established and examination reliability of the nozzle to shell weld area was improved. With the development of our 3-axis control system for PaR ISI-2 computer control system, the reliability of nozzle to shell weld area examination has been improved. The established technologies from the development and detailed analysis of existing control system, are expected to be applied to the similar control systems in nuclear power plants. (author). 12 refs., 4 tabs., 33 figs.

  8. Simple analysis of an External Vessel Cooling Thermosyphon for a Sodium-cooled Fast Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Choi, Jae Young; Jeong, Yong Hoon [Korea Advanced Institute of Science and Technology, Daejeon (Korea, Republic of); Song, Sub Lee [Handong Global University, Pohang (Korea, Republic of)

    2015-05-15

    KALIMER has three different DHR systems: two non-safety grade systems and one safety grade system. The non-safety grade systems are an IRACS (Intermediate Reactor Auxiliary Cooling System) and a steam/feedwater system. The safety grade system is a PDRC (Passive Decay Heat Removal Circuit). In case of the foreign reactor designs, ABTR (Advanced Burner Test Reactor) has a DRACS (Direct Reactor Auxiliary Cooling System), a PFBR (Indian Prototype Fast Breeder Reactor) has an SGDHRS (Safety Grade Decay Heat Removal System), and an EFR (European Fast Reactor) has DRC (Direct Reactor Cooling). Those designs have advantage on relatively high decay heat removal capacity. However, larger vessel size due to subsidiary in-vessel structure and possible accident propagation to reactor induced by sodium fire. In this paper, an ex-vessel thermosyphon design was proposed for the removal of decay heat for an iSFR. The proposed ex-vessel thermosyphon was designed to remove decay heat in both transient cases and BDBA cases, such as vessel failure. Proper working fluid was selected based on thermodynamic properties and chemical stability. Mercury was chosen as the working fluid, and SUS 314 was used for the corresponding structure material. Possible chemical reactions and adverse effects from using the thermosyphon were inherently eliminated by the system layout. A model for a high-temperature thermosyphon and numerical algorithms were used for the analysis. As a result of the simulation, the thermosyphon design was optimized, and it showed sufficient DHR performance to maintain core integrity.

  9. Reactor Pressure Vessel Fracture Analysis Capabilities in Grizzly

    Energy Technology Data Exchange (ETDEWEB)

    Spencer, Benjamin [Idaho National Lab. (INL), Idaho Falls, ID (United States); Backman, Marie [Univ. of Tennessee, Knoxville, TN (United States); Chakraborty, Pritam [Idaho National Lab. (INL), Idaho Falls, ID (United States); Hoffman, William [Univ. of Idaho, Moscow, ID (United States)

    2015-03-01

    Efforts have been underway to develop fracture mechanics capabilities in the Grizzly code to enable it to be used to perform deterministic fracture assessments of degraded reactor pressure vessels (RPVs). Development in prior years has resulted a capability to calculate -integrals. For this application, these are used to calculate stress intensity factors for cracks to be used in deterministic linear elastic fracture mechanics (LEFM) assessments of fracture in degraded RPVs. The -integral can only be used to evaluate stress intensity factors for axis-aligned flaws because it can only be used to obtain the stress intensity factor for pure Mode I loading. Off-axis flaws will be subjected to mixed-mode loading. For this reason, work has continued to expand the set of fracture mechanics capabilities to permit it to evaluate off-axis flaws. This report documents the following work to enhance Grizzly’s engineering fracture mechanics capabilities for RPVs: • Interaction Integral and -stress: To obtain mixed-mode stress intensity factors, a capability to evaluate interaction integrals for 2D or 3D flaws has been developed. A -stress evaluation capability has been developed to evaluate the constraint at crack tips in 2D or 3D. Initial verification testing of these capabilities is documented here. • Benchmarking for axis-aligned flaws: Grizzly’s capabilities to evaluate stress intensity factors for axis-aligned flaws have been benchmarked against calculations for the same conditions in FAVOR. • Off-axis flaw demonstration: The newly-developed interaction integral capabilities are demon- strated in an application to calculate the mixed-mode stress intensity factors for off-axis flaws. • Other code enhancements: Other enhancements to the thermomechanics capabilities that relate to the solution of the engineering RPV fracture problem are documented here.

  10. 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.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2001-03-01

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

  12. 10 CFR 50.66 - Requirements for thermal annealing of the reactor pressure vessel.

    Science.gov (United States)

    2010-01-01

    ... Requirements for thermal annealing of the reactor pressure vessel. (a) For those light water nuclear power... life of these components. (B) The effects of localized high temperatures on degradation of the concrete... thermal annealing or to operate the nuclear power reactor following the annealing must be identified....

  13. Development of Observation Techniques in Reactor Vessel of Experimental Fast Reactor Joyo

    Science.gov (United States)

    Takamatsu, Misao; Imaizumi, Kazuyuki; Nagai, Akinori; Sekine, Takashi; Maeda, Yukimoto

    In-Vessel Observations (IVO) techniques for Sodium cooled Fast Reactors (SFRs) are important in confirming its safety and integrity. And several IVO equipments for an SFR are developed. However, in order to secure the reliability of IVO techniques, it was necessary to demonstrate the performance under the actual reactor environment with high temperature, high radiation dose and remained sodium. During the investigation of an incident that occurred with Joyo, IVO using a standard Video Camera (VC) and a Radiation-Resistant Fiberscope (RRF) took place at (1) the top of the Sub-Assemblies (S/As) and the In-Vessel Storage rack (IVS), (2) the bottom face of the Upper Core Structure (UCS). A simple 6 m overhead view of each S/A, through the fuel handling or inspection holes etc, was photographed using a VC for making observations of the top of S/As and IVS. About 650 photographs were required to create a composite photograph of the top of the entire S/As and IVS, and a resolution was estimated to be approximately 1mm. In order to observe the bottom face of the UCS, a Remote Handling Device (RHD) equipped with RRFs (approximately 13 m long) was specifically developed for Joyo with a tip that could be inserted into the 70 mm gap between the top of the S/As and the bottom of the UCS. A total of about 35,000 photographs were needed for the full investigation. Regarding the resolution, the sodium flow regulating grid of 0.8mm in thickness could be discriminated. The performance of IVO equipments under the actual reactor environment was successfully confirmed. And the results provided useful information on incident investigations. In addition, fundamental findings and the experience gained during this study, which included the design of equipment, operating procedures, resolution, lighting adjustments, photograph composition and the durability of the RRF under radiation exposure, provided valuable insights into further improvements and verifications for IVO techniques to

  14. Nuclear reactor pressure vessel-specific flaw distribution development

    Energy Technology Data Exchange (ETDEWEB)

    Rosinski, S.T.

    1992-09-01

    Vessel integrity predictions performed through fracture mechanics analysis of a pressurized thermal shock event have been shown to be significantly sensitive to the overall flaw distribution input. It has also been shown that modem vessel in-service inspection (ISI) results can be used for development of vessel flaw distribution(s) that are more representative of US vessels. This paper describes the development and application of a methodology to analyze ISI data for the purpose of flaw distribution determination. The resultant methodology considers detection reliability, flaw sizing accuracy, and flaw detection threshold in its application. Application of the methodology was then demonstrated using four recently acquired US PWR vessel inspection data sets. Throughout the program, new insight was obtained into several key inspection performance and vessel integrity prediction practice issues that will impact future vessel integrity evaluation. For example, the potential application of a vessel-specific flaw distribution now provides at least one method by which a vessel-specific reference flaw size applicable to pressure-temperature limit curves determination can be estimated. This paper will discuss the development and application of the methodology and the impact to future vessel integrity analyses.

  15. Nuclear reactor pressure vessel-specific flaw distribution development

    Energy Technology Data Exchange (ETDEWEB)

    Rosinski, S.T.

    1992-01-01

    Vessel integrity predictions performed through fracture mechanics analysis of a pressurized thermal shock event have been shown to be significantly sensitive to the overall flaw distribution input. It has also been shown that modem vessel in-service inspection (ISI) results can be used for development of vessel flaw distribution(s) that are more representative of US vessels. This paper describes the development and application of a methodology to analyze ISI data for the purpose of flaw distribution determination. The resultant methodology considers detection reliability, flaw sizing accuracy, and flaw detection threshold in its application. Application of the methodology was then demonstrated using four recently acquired US PWR vessel inspection data sets. Throughout the program, new insight was obtained into several key inspection performance and vessel integrity prediction practice issues that will impact future vessel integrity evaluation. For example, the potential application of a vessel-specific flaw distribution now provides at least one method by which a vessel-specific reference flaw size applicable to pressure-temperature limit curves determination can be estimated. This paper will discuss the development and application of the methodology and the impact to future vessel integrity analyses.

  16. Device for cooling the main vessel of a fast fission nuclear reactor

    Energy Technology Data Exchange (ETDEWEB)

    Debru, M.

    1984-10-16

    The annular space delimited by the main vessel and an internal shell is in communication with the zone of the reactor vessel, in which the cold primary liquid is located. The annular space delimited by the shell and by an internal shell is in communication with the lower part of the core via tubes. Thus, the cold primary liquid is injected into the space where it circulates from bottom to top, and flows into the space, where it circulates from top to bottom while at the same time cooling the main vessel. The invention applies, in particular, to fast fission nuclear reactors cooled by liquid sodium.

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

    CERN Document Server

    American Society for Testing and Materials. Philadelphia

    2010-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

    Chang, S.J.

    1997-05-01

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

  19. Improving the neutronic characteristics of a boiling water reactor by using uranium zirconium hydride fuel instead of uranium dioxide fuel

    Energy Technology Data Exchange (ETDEWEB)

    Galahom, Ahmed Abdelghafar [Higher Technological Institute, Ramadan (Egypt)

    2016-06-15

    The present work discusses two different models of boiling water reactor (BWR) bundle to compare the neutronic characteristics of uranium dioxide (UO{sub 2}) and uranium zirconium hydride (UZrH{sub 1.6}) fuel. Each bundle consists of four assemblies. The BWR assembly fueled with UO{sub 2} contains 8 × 8 fuel rods while that fueled with UZrH{sub 1.6} contains 9 × 9 fuel rods. The Monte Carlo N-Particle Transport code, based on the Mont Carlo method, is used to design three dimensional models for BWR fuel bundles at typical operating temperatures and pressure conditions. These models are used to determine the multiplication factor, pin-by-pin power distribution, axial power distribution, thermal neutron flux distribution, and axial thermal neutron flux. The moderator and coolant (water) are permitted to boil within the BWR core forming steam bubbles, so it is important to calculate the reactivity effect of voiding at different values. It is found that the hydride fuel bundle design can be simplified by eliminating water rods and replacing the control blade with control rods. UZrH{sub 1.6} fuel improves the performance of the BWR in different ways such as increasing the energy extracted per fuel assembly, reducing the uranium ore, and reducing the plutonium accumulated in the BWR through burnup.

  20. 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)

  1. R- AND P- REACTOR VESSEL IN-SITU DECOMISSIONING VISUALIZATION

    Energy Technology Data Exchange (ETDEWEB)

    Vrettos, N.; Bobbitt, J.; Howard, M.

    2010-06-07

    The R- & P- Reactor facilities were constructed in the early 1950's in response to Cold War efforts. The mission of the facilities was to produce materials for use in the nation's nuclear weapons stockpile. R-Reactor was removed from service in 1964 when President Johnson announced a slowdown of he nuclear arms race. PReactor continued operation until 1988 until the facility was taken off-line to modernize the facility with new safeguards. Efforts to restart the reactor ended in 1990 at the end of the Cold War. Both facilities have sat idle since their closure and have been identified as the first two reactors for closure at SRS.

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

    CERN Document Server

    American Society for Testing and Materials. Philadelphia

    2003-01-01

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

  3. 3D TRANSIENT COUPLED THERMO-ELASTIC-PLASTIC CONTACT SEALING ANALYSIS OF REACTOR PRESSURE VESSEL

    Institute of Scientific and Technical Information of China (English)

    Du Xuesong; Li Runfang; Lin Tengjiao

    2005-01-01

    Sealing analysis of sealing system in reactor pressure vessels is relevant with multiple nonlinear coupled-field effects, so even large-scale commercial finite element software cannot finish the complicated analysis. A fmite element method of 3D transient coupled thermo-elastic-plastic contact sealing analysis for reactor pressure vessels is presented, in which the surface nonlinearity,material nonlinearity, transient heat transfer nonlinearity and multiple coupled effect are taken into account and the sealing equation is coupling solved in iterative procedure. At the same time, a computational analysis program is developed, which is applied in the sealing analysis of experimental reactor pressure vessel, and the numerical results are in good coincidence with the experimental results. This program is also successful in analyzing the practical problem in engineering.

  4. Reactor Pressure Vessel Temperature Analysis for Prismatic and Pebble-Bed VHTR Designs

    Energy Technology Data Exchange (ETDEWEB)

    H. D. Gougar; C. B. Davis

    2006-04-01

    Analyses were performed to determine maximum temperatures in the reactor pressure vessel for two potential Very-High Temperature Reactor (VHTR) designs during normal operation and during a depressurized conduction cooldown accident. The purpose of the analyses was to aid in the determination of appropriate reactor vessel materials for the VHTR. The designs evaluated utilized both prismatic and pebble-bed cores that generated 600 MW of thermal power. Calculations were performed for fluid outlet temperatures of 900 and 950 °C, corresponding to the expected range for the VHTR. The analyses were performed using the RELAP5-3D and PEBBED-THERMIX computer codes. Results of the calculations were compared with preliminary temperature limits derived from the ASME pressure vessel code.

  5. Proceedings of the DOE/SNL/EPRI sponsored Reactor Pressure Vessel Thermal Annealing Workshop. Volume 1

    Energy Technology Data Exchange (ETDEWEB)

    Rosinski, S.T. [ed.] [Sandia National Labs., Albuquerque, NM (United States); Carter, R.G. [ed.] [Electric Power Research Institute, Charlotte, NC (United States)

    1994-09-01

    The purpose of the Reactor Pressure vessel Thermal Annealing Workshop was to provide a forum for US utilities and interested parties to discuss relevant experience and issues and identify potential solutions/approaches related to: An understanding of the potential benefits of thermal annealing for US commercial reactors; on-going technical research activities; technical aspects of a generic, full-scale, in-place vessel annealing demonstration; and the impact of economic, regulatory, and technical issues on the application of thermalannealingtechnology to US plants. Experts from the international nuclear reactor community were brought together to discuss issues regarding application of thermal annealing technology in the US and identify the steps necessary to commercialize this technology for US reactors. These proceedings contain all presentation materials discussed during the Workshop. Selected papers are indexed separately for inclusion in the Energy Science and Technology Database.

  6. Proceedings of the DOE/SNL/EPRI sponsored Reactor Pressure Vessel Thermal Annealing Workshop. Volume 2

    Energy Technology Data Exchange (ETDEWEB)

    Rosinski, S.T. [ed.] [Sandia National Labs., Albuquerque, NM (United States); Carter, R.G. [ed.] [Electric Power Research Institute, Charlotte, NC (United States)

    1994-09-01

    The purpose of the Reactor Pressure Vessel Thermal Annealing Workshop was to provide a forum for US utilities and interested parties to discuss relevant experience and issues and identify potential solutions/approaches related to: (1) an understanding of the potential benefits of thermal annealing for US commercial reactors; (2) on-going technical research activities; (3) technical aspects of a generic, full-scale, in-place vessel annealing demonstration; and (4) the impact of economic, regulatory, and technical issues on the application of thermal annealing technology to US plants. Experts from the international nuclear reactor community were brought together to discuss issues regarding application of thermal annealing technology in the US and identify the steps necessary to commercialize this technology for US reactors. These proceedings contain all presentation materials discussed during the Workshop. This document, Volume 2, contains sections 10 through 13, Individual papers have been cataloged separately.

  7. Retrospective Dosimetry of Vver 440 Reactor Pressure Vessel at the 3RD Unit of Dukovany Npp

    Science.gov (United States)

    Marek, M.; Viererbl, L.; Sus, F.; Klupak, V.; Rataj, J.; Hogel, J.

    2009-08-01

    Reactor pressure vessel (RPV) residual lifetime of the Czech VVER-440 is currently monitored under Surveillance Specimens Programs (SSP) focused on reactor pressure vessel materials. Neutron fluence in the samples and its distribution in the RPV are determined by a combination of calculation results and the experimental data coming from the reactor dosimetry measurements both in the specimen containers and in the reactor cavity. The direct experimental assessment of the neutron flux density incident onto RPV and neutron fluence for the entire period of nuclear power plant unit operation can be based on the evaluation of the samples taken from the inner RPV cladding. The Retrospective Dosimetry was also used at Dukovany NPP at its 3rd unit after the 18th cycle. The paper describes methodology, experimental setup for sample extraction, measurement of activities, and the determination of the neutron flux and fluence averaged over the samples.

  8. Mechanical properties of reactor pressure vessel steels studied by static and dynamic torsion tests

    Science.gov (United States)

    Munier, A.; Maamouri, M.; Schaller, R.; Mercier, O.

    1993-06-01

    Internal friction measurements and torsional plastic deformation tests have been performed in reactor pressure vessel steels (unirradiated, irradiated and irradiated/annealed specimens). The results of these experiments have been interpreted with help of transmission electron microscopy observations (conventional and in situ). It is shown how the interactions between screw dislocations and obstacles (Peierls valleys, impurities and precipitates) could explain the low temperature hardening and the irradiation embrittlement of ferritic steels. In addition, it appears that the nondestructive internal friction technique could be used advantageously to follow the evolution of the material properties under irradiation, as for instance the irradiation embrittlement of the reactor pressure vessel steels.

  9. Dosimetry analyses of the Ringhals 3 and 4 reactor pressure vessels

    Energy Technology Data Exchange (ETDEWEB)

    Kulesza, J.A.; Fero, A.H. [Westinghouse Electric Company, Cranberry Township, PA 16066 (United States); Rouden, J.; Green, E.L. [Vattenfall/Ringhals AB, 432 85 Vaeroebacka (Sweden)

    2011-07-01

    A comprehensive series of neutron dosimetry measurements consisting of surveillance capsules, reactor pressure vessel cladding samples, and ex-vessel neutron dosimetry has been analyzed and compared to the results of three-dimensional, cycle-specific neutron transport calculations for the Ringhals Unit 3 and Unit 4 reactors in Sweden. The comparisons show excellent agreement between calculations and measurements. The measurements also demonstrate that it is possible to perform retrospective dosimetry measurements using the {sup 93}Nb (n,n') {sup 93m}Nb reaction on samples of 18-8 austenitic stainless steel with only trace amounts of elemental niobium. (authors)

  10. Research Progress of Irradiation Embrittlement Behavior and Prediction Technology of Reactor Pressure Vessel Steel

    Institute of Scientific and Technical Information of China (English)

    YANG; Wen; TONG; Zhen-feng; NING; Guang-sheng; ZHANG; Chang-yi; BAI; Bing

    2015-01-01

    The reactor pressure vessel(RPV)is the core of the most important equipment in pressurized water reactor,and is the key equipment that cannot be replaced in nuclear power plant.The service life of RPV determines the use of nuclear power plant,and directly affects the safety and economy of nuclear power plant.Because of high temperature,high pressure and high-energy

  11. LOFT reactor vessel 290/sup 0/ downcomer stalk instrument penetration flange stress analysis

    Energy Technology Data Exchange (ETDEWEB)

    Finicle, D.P.

    1978-06-06

    The LOFT Reactor Vessel 290/sup 0/ Downcomer Stalk Instrument Penetration Flange Stress Analysis has been completed using normal operational and blowdown loading. A linear elastic analysis was completed using simplified hand analysis techniques. The analysis was in accordance with the 1977 ASME Boiler and Pressure Vessel Code, Section III, for a Class 1 component. Loading included internal pressure, bolt preload, and thermal gradients due to normal operating and blowdown.

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

    Energy Technology Data Exchange (ETDEWEB)

    Boing, L.E.; Henley, D.R. (Argonne National Lab., IL (USA)); Manion, W.J.; Gordon, J.W. (Nuclear Energy Services, Inc., Danbury, CT (USA))

    1989-12-01

    Metal cutting techniques that can be used to segment the reactor pressure vessel of the Experimental Boiling Water Reactor (EBWR) at Argonne National Laboratory (ANL) have been evaluated by Nuclear Energy Services. Twelve cutting technologies are described in terms of their ability to perform the required task, their performance characteristics, environmental and radiological impacts, and cost and schedule considerations. Specific recommendations regarding which technology should ultimately be used by ANL are included. The selection of a cutting method was the responsibility of the decommissioning staff at ANL, who included a relative weighting of the parameters described in this document in their evaluation process. 73 refs., 26 figs., 69 tabs.

  13. Boiling water reactor in a prestressed reinforced concrete vessel for an atomic central heating-and-power plant

    Energy Technology Data Exchange (ETDEWEB)

    Tokarev, Yu.I.; Sokolov, I.N.; Skvortsov, S.A.; Sidorov, A.M.; Krauze, L.V.

    1978-04-01

    The possibility of using a boiling water reactor in a prestressed reinforced concrete vessel for an atomic central heating-and-power plant (CHPP) was considered, with design features of the reactor intended for a two-purpose plant. A prestressed reinforced concrete vessel and integral arrangement of the primary circuit ensured reliability of the atomic CHPP using various CHPP flowsheets.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2007-01-15

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2006-11-15

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

  16. Final report of the 1st ex-vessel neutron dosimetry installation and evaluations for Yonggwang unit 2 reactor pressure vessel

    Energy Technology Data Exchange (ETDEWEB)

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

    2006-09-15

    This report describes a neutron fluence assessment performed for the Yonggwang unit 2 pressure vessel beltline region based on the guidance specified in regulatory guide 1.190. In this assessment, maximum fast neutron exposures expressed in terms of fast neutron fluence (E>1 MeV) and iron atom displacements (dpa) were established for the beltline region of the pressure vessel. During cycle 15 of reactor operation, an ex-vessel neutron dosimetry program was instituted at Yonggwang unit 2 to provide continuous monitoring of the beltline region of the reactor vessel. The use of the ex-vessel neutron dosimetry program coupled with available surveillance capsule measurements provides a plant specific data base that enables the evaluation of the vessel exposure and the uncertainty associated with that exposure over the service life of the unit. Ex-vessel neutron dosimetry has been evaluated at the conclusion of cycle 15.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2006-11-15

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2007-02-15

    This report describes a neutron fluence assessment performed for the Kori Unit 1 pressure vessel beltline region based on the guidance specified in Regulatory Guide 1.190. In this assessment, maximum fast neutron exposures expressed in terms of fast neutron fluence (E>1 MeV) and iron atom displacements (dpa) were established for the beltline region of the pressure vessel. After Cycle 22 of reactor operation, 2nd Ex-Vessel Neutron Dosimetry Program was instituted at Kori Unit 1 to provide continuous monitoring of the beltline region of the reactor vessel. The use of the Ex-Vessel Neutron Dosimetry Program coupled with available surveillance capsule measurements provides a plant specific data base that enables the evaluation of the vessel exposure and the uncertainty associated with that exposure over the service life of the unit. Ex-Vessel Neutron Dosimetry has been evaluated at the conclusion of Cycle 23.

  19. Final report for the 3rd Ex-Vessel Neutron Dosimetry Installations and Evaluations for Kori Unit 1 Reactor Pressure Vessel

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Byoung Chul; Yoo, Choon Sung; Lee, Sam Lai (and others)

    2008-03-15

    This report describes a neutron fluence assessment performed for the Kori Unit 1 pressure vessel belt line region based on the guidance specified in Regulatory Guide 1.190. In this assessment, maximum fast neutron exposures expressed in terms of fast neutron fluence (E>1 MeV) and iron atom displacements (dpa) were established for the belt line region of the pressure vessel. After Cycle 23 of reactor operation, 3rd Ex-Vessel Neutron Dosimetry Program was instituted at Kori Unit 1 to provide continuous monitoring of the belt line region of the reactor vessel. The use of the Ex-Vessel Neutron Dosimetry Program coupled with available surveillance capsule measurements provides a plant specific data base that enables the evaluation of the vessel exposure and the uncertainty associated with that exposure over the service life of the unit. Ex-Vessel Neutron Dosimetry has been evaluated at the conclusion of Cycle 24.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2008-01-15

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2008-01-15

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

  2. 77 FR 16270 - Updated Aging Management Criteria for Reactor Vessel Internal Components of Pressurized Water...

    Science.gov (United States)

    2012-03-20

    ... COMMISSION Updated Aging Management Criteria for Reactor Vessel Internal Components of Pressurized Water... license renewal interim staff guidance (LR-ISG), LR-ISG-2011-04, ``Updated Aging Management Criteria for... Aging Lessons Learned (GALL) Report for the aging management of stainless steel structures...

  3. Simulation of the Lower Head Boiling Water Reactor Vessel in a Severe Accident

    Directory of Open Access Journals (Sweden)

    Alejandro Nuñez-Carrera

    2012-01-01

    Full Text Available The objective of this paper is the simulation and analysis of the BoilingWater Reactor (BWR lower head during a severe accident. The COUPLE computer code was used in this work to model the heatup of the reactor core material that slumps in the lower head of the reactor pressure vessel. The prediction of the lower head failure is an important issue in the severe accidents field, due to the accident progression and the radiological consequences that are completely different with or without the failure of the Reactor Pressure Vessel (RPV. The release of molten material to the primary containment and the possibility of steam explosion may produce the failure of the primary containment with high radiological consequences. Then, it is important to have a detailed model in order to predict the behavior of the reactor vessel lower head in a severe accident. In this paper, a hypothetical simulation of a Loss of Coolant Accident (LOCA with simultaneous loss of off-site power and without injection of cooling water is presented with the proposal to evaluate the temperature distribution and heatup of the lower part of the RPV. The SCDAPSIM/RELAP5 3.2 code was used to build the BWR model and conduct the numerical simulation.

  4. Aging of reactor vessels in LWR type reactors; Envejecimiento de la vasija y de los internos del nuclear de los reactores tipo LWR

    Energy Technology Data Exchange (ETDEWEB)

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

    2004-07-01

    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.

  5. Dosimetry assessments for the reactor pressure vessel and core barrel in UK PWR plant

    Energy Technology Data Exchange (ETDEWEB)

    Thornton, D.A.; Allen, D.A.; Huggon, A.P.; Picton, D.J.; Robinson, A.T.; Steadman, R.J. [Serco, Rutherford House, Quedgeley, Gloucester, Gl2 4NF (United Kingdom); Seren, T.; Lipponen, M.; Kekki, T. [VTT, Technical Research Centre of Finland, Otakaari 3 K, P.O. BOX 1000, Espoo, FI-02044 (Finland)

    2011-07-01

    Specimens for the Sizewell B reactor pressure vessel (RPV) inservice steels surveillance program are irradiated inside eight capsules located within the reactor pressure vessel and loaded prior to commissioning. The periodic removal of these capsules and testing of their contents provides material properties data at intervals during the lifetime of the plant. Neutron activation measurements and radiation transport calculations play an essential role in assessing the neutron exposure of the specimens and RPV. Following the most recent withdrawal, seven capsules have now been removed covering nine cycles of reactor operation. This paper summarizes the dosimetry results of the Sizewell B surveillance program obtained to date. In addition to an overview of the calculational methodology it includes a review of the measurements. Finally, it describes an extension of the methodology to provide dosimetry recommendations for the core barrel and briefly discusses the results that were obtained. (authors)

  6. Severe accident improvements for Carem-25 to arrest reactor vessel meltdown sequences

    Energy Technology Data Exchange (ETDEWEB)

    Poier Baez, L.E.; Nunez Mac Leod, J.E.; Baron, J.H. [Cuyo National University, Engineering Faculty, Mendoza (Argentina)

    2001-07-01

    Given an accident sequence, that leads to sustained uncovering of the core, the progression of core damage involves several complex phenomena. The progression of these phenomena can lead to a breach of the reactor vessel followed by the discharge of molten core materials to the containment. Advanced nuclear reactor designs, such as the CAREM reactor, include several improvements related to safety issues either enhancing the passive safety functions or allowing plant operators more time to undertake different management actions against radioactive releases to the environment. In the development of the nuclear power plant CAREM, the possibility of including a passive metallic in-vessel container in its design is being considered, to arrest the reactor pressure vessel meltdown sequence during a core damaging event, and thereof prevent its failure. The paper comprises the first analyses, via numerical simulation, for the conceptual design of such a container type; furthermore, the paper addresses simulation model characteristics helping to establish geometrical dimensions, materials and container compatibility with power plant engineering features. The paper also presents the first model developed to analyze the complex relocation phenomena in the core of CAREM during a severe accident sequence caused by a loss of coolant. The PC version of MELCOR 1.8.4 code has been used to predict the transient behavior of core parameters. MELCOR is a fully integrated relatively fast running code that models the progression of accidents in light water reactor power plants. This paper presents reactor variables behavior during the first hours of the event being studied, giving preliminary conclusions about the use and capability of a metallic in-vessel core catcher. (authors)

  7. Severe accident improvements for Carem-25 to arrest reactor vessel meltdown sequences

    Energy Technology Data Exchange (ETDEWEB)

    Poier Baez, L.E.; Nunez Mac Leod, J.E.; Baron, J.H. [Cuyo National University, Engineering Faculty, Mendoza (Argentina)

    2001-07-01

    Given an accident sequence, that leads to sustained uncovering of the core, the progression of core damage involves several complex phenomena. The progression of these phenomena can lead to a breach of the reactor vessel followed by the discharge of molten core materials to the containment. Advanced nuclear reactor designs, such as the CAREM reactor, include several improvements related to safety issues either enhancing the passive safety functions or allowing plant operators more time to undertake different management actions against radioactive releases to the environment. In the development of the nuclear power plant CAREM, the possibility of including a passive metallic in-vessel container in its design is being considered, to arrest the reactor pressure vessel meltdown sequence during a core damaging event, and thereof prevent its failure. The paper comprises the first analyses, via numerical simulation, for the conceptual design of such a container type; furthermore, the paper addresses simulation model characteristics helping to establish geometrical dimensions, materials and container compatibility with power plant engineering features. The paper also presents the first model developed to analyze the complex relocation phenomena in the core of CAREM during a severe accident sequence caused by a loss of coolant. The PC version of MELCOR 1.8.4 code has been used to predict the transient behavior of core parameters. MELCOR is a fully integrated relatively fast running code that models the progression of accidents in light water reactor power plants. This paper presents reactor variables behavior during the first hours of the event being studied, giving preliminary conclusions about the use and capability of a metallic in-vessel core catcher. (authors)

  8. A conceptual design of a low resistance vacuum vessel for the Steady State Tokamak Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Suzuki, Yutaka; Yamada, Masao; Tomita, Mitsuru (Mitsubishi Fusion Center, Tokyo (Japan)); Kikuchi, Mitsuru; Nishio, Satoshi; Seki, Yasushi (Japan Atomic Energy Research Inst., Naka, Ibaraki (Japan))

    1991-12-01

    A design study on the vacuum vessel of the Steady State Tokamak Reactor has been performed in order to provide a realistic structural concept for a fusion reactor. The vacuum vessel and shield are integrated to form a double-thin-wall structure filled with stainless steel and water resulting in a low one-turn electric resistance of {proportional to}4 {mu}{Omega} without insulating breaks or bellows. The reinforcement plates are welded between the inner and outer skins of the double-thin-wall structure, and shielding units are installed in every chamber with electrical insulation from these skins and plates. As a result, the requirements for the vacuum vessel can be realized by this simple structure alone. Transient electromagnetic and structural analysis has been performed for a three-dimensional shell model in the plasma disruption condition of plasma current 12 MA and current decay time 20 ms. An eddy current, about 95% of plasma current, is induced on the vacuum vessel, and a maximum magnetic pressure {proportional to}5.8 MPa is caused by the coupling with the toroidal field. The maximum stress intensity for the magnetic pressure is about 216 MPa. This low resistance vacuum vessel is extremely effective in shielding the change of the magnetic field in the superconducting toroidal and poloidal field coils during a plasma disruption. In summary, the feasibility and features of this new type of vacuum vessel concept have been shown in this study. (orig.).

  9. Development and application of an LWR reactor pressure vessel-specific flaw distribution

    Energy Technology Data Exchange (ETDEWEB)

    Rosinski, S.T. (Sandia National Labs., Albuquerque, NM (United States)); Kennedy, E.L.; Foulds, J.R. (Failure Analysis Associates, Inc., Menlo Park, CA (United States))

    1991-01-01

    Previous efforts by the US Department of Energy have shown that the PWR reactor vessel integrity predictions performed through probabilistic fracture mechanics analysis for a pressurized thermal shock event are significantly sensitive to the overall flaw distribution input. It has also been shown that modern vessel in-service inspection (ISI) results can be used for development of vessel flaw distribution(s) that are more representative of US vessels. This paper describes the development and application of a methodology to analyze ISI data for the purpose of flaw distribution determination. The resultant methodology considers detection reliability, flaw sizing accuracy, and flaw detection threshold in its application. Application of the methodology was then demonstrated using four recently acquired US PWR vessel inspection data sets. The methodology helped provide original insight into several key inspection performance and vessel integrity prediction practice issues that will impact future vessel integrity evaluation. This paper briefly discusses the development and application of the methodology and the impact to future vessel integrity analyses.

  10. Development and application of an LWR reactor pressure vessel-specific flaw distribution

    Energy Technology Data Exchange (ETDEWEB)

    Rosinski, S.T. [Sandia National Labs., Albuquerque, NM (United States); Kennedy, E.L.; Foulds, J.R. [Failure Analysis Associates, Inc., Menlo Park, CA (United States)

    1991-12-31

    Previous efforts by the US Department of Energy have shown that the PWR reactor vessel integrity predictions performed through probabilistic fracture mechanics analysis for a pressurized thermal shock event are significantly sensitive to the overall flaw distribution input. It has also been shown that modern vessel in-service inspection (ISI) results can be used for development of vessel flaw distribution(s) that are more representative of US vessels. This paper describes the development and application of a methodology to analyze ISI data for the purpose of flaw distribution determination. The resultant methodology considers detection reliability, flaw sizing accuracy, and flaw detection threshold in its application. Application of the methodology was then demonstrated using four recently acquired US PWR vessel inspection data sets. The methodology helped provide original insight into several key inspection performance and vessel integrity prediction practice issues that will impact future vessel integrity evaluation. This paper briefly discusses the development and application of the methodology and the impact to future vessel integrity analyses.

  11. KTA 3401. 1. Reactor safety vessel of steel

    Energy Technology Data Exchange (ETDEWEB)

    1983-04-13

    The text of the standard has been prepared by order of the Nuclear Committee of the Working Group on Pressure Vessels with the ''Verein Deutscher Eisenhuettenleute (VDEhL)'' acting as main contractor. This standard replaces the standard KTA 3401.1, edition 6/80. As against edition 6/80 the text of the standard has been editorially treated, in particular for adaptation to the newly included annex A: ''Material characteristics''. Steels: 15MnNi63 (DIN-1.6210); 40NiCrMo84 (DIN-1,6562); 26NiCrMo146 (DIN-1.6958); 20NiCrMo145 (DIN-1.6772); 34CrMo4 (DIN-1.7220); 42CrMo4 (DIN-1.7225); C45 (DIN-1.0503).

  12. ASSESSMENT OF THE POTENTIAL FOR HYDROGEN GENERATION DURING DEACTIVATION AND DECOMMISSIONING OF REACTOR VESSELS AT THE SAVANNAH RIVER SITE

    Energy Technology Data Exchange (ETDEWEB)

    Wiersma, B.; Serrato, M.; Langton, C.

    2010-11-10

    The R- and P-reactor vessels at the Savannah River Site (SRS) are being prepared for deactivation and decommissioning (D&D). D&D activities will consist primarily of physically isolating and stabilizing the reactor vessel by filling it with a grout material. The reactor vessels contain aluminum alloy materials, which pose a concern in that aluminum corrodes rapidly when it comes in contact with the alkaline grout. A product of the corrosion reaction is hydrogen gas and therefore potential flammability issues were assessed. A model was developed to calculate the hydrogen generation rate as the reactor is being filled with the grout material. Three options existed for the type of grout material for D&D of the reactor vessels. The grout formulation options included ceramicrete (pH 6-8), a calcium aluminate sulfate (CAS) based cement (pH 10), or Portland cement grout (pH 12.4). Corrosion data for aluminum in concrete were utilized as input for the model. The calculations considered such factors as the surface area of the aluminum components, the open cross-sectional area of the reactor vessel, the rate at which the grout is added to the reactor vessel, and temperature. Given the hydrogen generation rate, the hydrogen concentration in the vapor space of the reactor vessel above the grout was calculated. This concentration was compared to the lower flammability limit for hydrogen. The assessment concluded that either ceramicrete or the CAS grout may be used to safely grout the P-reactor vessel. The risk of accumulation of a flammable mixture of hydrogen between the grout-air interface and the top of the reactor is very low. Portland cement grout, on the other hand, for the same range of process parameters did not provide a margin of safety against the accumulation of flammable gas in the reactor vessel during grouting operations in the P-reactor vessel. Therefore, it was recommended that this grout not be utilized for this task. On the other hand, the R-reactor vessel

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2005-01-01

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

  14. Irradiation embrittlement of reactor pressure vessel steel outside the astm specification A508 CL2

    Science.gov (United States)

    Pachur, D.; Krawczynski, S. J.; Derz, H.; Pott, G.

    1990-04-01

    Radiation embrittlement of reactor pressure vessel steels is of considerable significance for safety engineering. Steel manufacturers must therefore comply with specifications defined by national design codes. The extent to which a steel deviating from the specification is influenced by irradiation is being examined under the German Research Programme on the Integrity of Reactor Components. Charpy-V specimens were taken from a forged steel block longitudinally and vertically to the direction of main deformation and irradiated in the FRJ-1 research reactor at a temperature of 288 °C corresponding to the operating temperature of power reactors. The neutron fluences obtained ranged between 0.8 × 10 19 and 8 × 10 19n/ cm2. Instrumented pendulum impact tests have been evaluated and the load signals measured were analysed, fitting and calculating transition temperature curves and trend curves.

  15. Flow boiling CHF enhancement in an external reactor vessel cooling (ERVC) channel using graphene oxide nanofluid

    Energy Technology Data Exchange (ETDEWEB)

    Park, Seong Dae; Bang, In Cheol, E-mail: icbang@unist.ac.kr

    2013-12-15

    Highlights: • We investigate CHF limits of graphene oxide nanofluid for IVR-ERVC. • Graphene oxide nanofluid enhanced CHF up to about 20%. • CHF enhancement can be explained by the improved thermal activity. - Abstract: External reactor vessel cooling for in-vessel retention of corium is an important concept to mitigate the consequences of a severe accident by flooding the reactor cavity. Although this system has some merits, it is restricted by the capacity of heat removal through the nucleate boiling on the outer surface of the reactor. In this study, the graphene oxide (GO) nanofluid at 0.0001 vol% was used to enhance the critical heat flux (CHF). The CHF tests were conducted with a closed-loop facility. Test section simulated the reactor vessel of APR-1400 with a small scale. The test results show about ∼20% enhancement of CHF at 50 and 100 kg/m{sup 2} s under a 10 K subcooling condition. It means that the additional thermal margin could be acquired by just adding the GO nanoparticles to the flooding water without severe economic concerns. It is also found that this CHF enhancement is caused by coating the graphene oxide nanoparticles on the heated surface. However, the sessile drop tests on the coated heater surface show that the wettability of GO coated surface is not improved. The results of IR thermography show that one of the promising reasons is the change of thermal activity due to the coated GO nanoparticles on the heated surface.

  16. Estimation of post-buckling fatigue damage for LMFBR reactor vessel under seismic load

    Energy Technology Data Exchange (ETDEWEB)

    Ogiso, S.; Sasaki, T.; Oooka, Y. [Kawasaki Heavy Industries, Ltd., Tokyo (Japan). Nuclear Systems Div.; Nakamura, H. [Central Research Inst. of Electric Power Industry, Chiba (Japan)

    1995-12-31

    Estimation of fatigue damage caused by buckling deformation is important to evaluate safety margin in a seismic buckling design criterion for LMFBR reactor vessels, in addition to limiting the buckling strength. An advanced buckling design guideline draft including the seismic margin criterion has been proposed under the sponsorship of MITI to date. An ultimate state in this criterion was defined as the condition that the maximum global displacement {delta}{sub max} reaches a critical displacement {delta}{sub u}. The authors have previously proposed an estimation method of the fatigue damage based on the post buckling fatigue tests 304 s.s. cylinders at room temperature. However, adoption of a modified 316 s.s named 316FR s.s is under development as the material of reactor vessel of the updated design of the Demonstration Fast Breeder Reactor. The buckling tests with 316FR s.s cylinders were performed under high temperature to obtain the skeleton curve of the relation between load and displacement. And the buckling behaviors under the cyclic loading were compared with those of 304 s.s. Objectives of the present study are: to apply the proposed estimation method to a reactor vessel made of 316FR s.s., and clarify the correlation between {delta}{sub max} and fatigue failure; to verify structural soundness of the ultimate state derived from the seismic margin criterion against the fatigue failure due to the buckling deformation. (author). 7 refs., 12 figs., 1 tab.

  17. Correlation between radiation damage and magnetic properties in reactor vessel steels

    Energy Technology Data Exchange (ETDEWEB)

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

    2014-02-01

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

  18. MAGNESIUM MONO POTASSIUM PHOSPHATE GROUT FOR P-REACTOR VESSEL IN-SITU DECOMISSIONING

    Energy Technology Data Exchange (ETDEWEB)

    Langton, C.; Stefanko, D.

    2011-01-05

    The objective of this report is to document laboratory testing of magnesium mono potassium phosphate grouts for P-Reactor vessel in-situ decommissioning. Magnesium mono potassium phosphate cement-based grout was identified as candidate material for filling (physically stabilizing) the 105-P Reactor vessel (RV) because it is less alkaline than portland cement-based grout (pH of about 12.4). A less alkaline material ({<=} 10.5) was desired to address a potential materials compatibility issue caused by corrosion of aluminum metal in highly alkaline environments such as that encountered in portland cement grouts. Information concerning access points into the P-Reactor vessel and amount of aluminum metal in the vessel is provided elsewhere. Fresh and cured properties were measured for: (1) commercially blended magnesium mono potassium phosphate packaged grouts, (2) commercially available binders blended with inert fillers at SRNL, (3) grouts prepared from technical grade MgO and KH{sub 2}PO{sub 4} and inert fillers (quartz sands, Class F fly ash), and (4) Ceramicrete{reg_sign} magnesium mono potassium phosphate-based grouts prepared at Argonne National Laboratory. Boric acid was evaluated as a set retarder in the magnesium mono potassium phosphate mixes.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2001-03-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2000-10-01

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

  1. Prediction of the reactor vessel water level using fuzzy neural networks in severe accident circumstance of NPPs

    Energy Technology Data Exchange (ETDEWEB)

    Park, Soon Ho; Kim, Dae Seop; Kim, Jae Hwan; Na, Man Gyun [Dept. of Nuclear Engineering, Chosun University, Gwangju (Korea, Republic of)

    2014-06-15

    Safety-related parameters are very important for confirming the status of a nuclear power plant. In particular, the reactor vessel water level has a direct impact on the safety fortress by confirming reactor core cooling. In this study, the reactor vessel water level under the condition of a severe accident, where the water level could not be measured, was predicted using a fuzzy neural network (FNN). The prediction model was developed using training data, and validated using independent test data. The data was generated from simulations of the optimized power reactor 1000 (OPR1000) using MAAP4 code. The informative data for training the FNN model was selected using the subtractive clustering method. The prediction performance of the reactor vessel water level was quite satisfactory, but a few large errors were occasionally observed. To check the effect of instrument errors, the prediction model was verified using data containing artificially added errors. The developed FNN model was sufficiently accurate to be used to predict the reactor vessel water level in severe accident situations where the integrity of the reactor vessel water level sensor is compromised. Furthermore, if the developed FNN model can be optimized using a variety of data, it should be possible to predict the reactor vessel water level precisely.

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

    Energy Technology Data Exchange (ETDEWEB)

    Saikusa, Akio [Secretariat of Nuclear Safety Commission, Tokyo (Japan); Nakagawa, Shigeaki; Fujimoto, Nozomu; Tachibana, Yukio; Iyoku, Tatsuo [Japan Atomic Energy Research Inst., Oarai, Ibaraki (Japan). Oarai Research Establishment

    2003-02-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Jong Po; Park, C. H.; Kim, H. T.; Noh, H. C.; Lee, J. M.; Kim, C. K.; Um, B. G. [Research Institute of KAITEC, Seoul (Korea)

    2002-03-01

    The objective of this project is to develop an automated ultrasonic data acquisition and data analysis system to examine heavy vessel welds. In order to examine nuclear vessel welds including reactor pressure vessel(RPV), huge amount of ultrasonic data from 6 channels should be able to be on-line processed. In addition, ultrasonic transducer scanning device should be remotely controlled, because working place is high radiation area. This kind of an automated ultrasonic testing equipment has not been developed domestically yet. In order to develop an automated ultrasonic testing system, RPV ultrasonic testing equipments developed in foreign countries were investigated and the capability of high speed ultrasonic signal processing hardwares was analyzed. In this study, ultrasonic signal processing system was designed. And also, ultrasonic data acquisition software was developed. The new systems were tested on the RPV welds of Ulchin Unit 6 to confirm their functions and capabilities. They worked very well as designed and the tests were successfully completed. 13 refs., 34 figs., 11 tabs. (Author)

  4. Collaborative investigations of in-service irradiated material from the Japan Power Demonstration Reactor pressure vessel

    Energy Technology Data Exchange (ETDEWEB)

    Corwin, W.R.; Broadhead, B.L. [Oak Ridge National Lab., TN (United States); Suzuki, M.; Kohsaka, A. [Japan Atomic Energy Research Institute, Tokai (Japan)

    1997-02-01

    There is a need to validate the results of irradiation effects research by the examination of material taken directly from the wall of a pressure vessel that has been irradiated during normal service. Just such an evaluation is currently being conducted on material from the wall of the pressure vessel from the Japan Power Demonstration Reactor (JPDR). The research is being jointly performed at the Tokai Research Establishment of the Japan Atomic Energy Research Institute (JAERI) and by the Nuclear Regulatory Commission (NRC)-funded Heavy-Section Steel Irradiation Program at the Oak Ridge National Laboratory (ORNL).

  5. Creep crack growth in a reactor pressure vessel steel at 360 deg C

    Energy Technology Data Exchange (ETDEWEB)

    Rui Wu; Seitisleam, F.; Sandstroem, R. [Swedish Institute for Metals Research, Stockholm (Sweden)

    1998-12-31

    Plain creep (PC) and creep crack growth (CCG) tests at 360 deg C and post metallography were carried out on a low alloy reactor pressure vessel steel (ASTM A508 class 2) with different microstructures. Lives for the CCG tests were shorter than those for the PC tests and this is more pronounced for simulated heat affected zone microstructure than for the parent metal at longer lives. For the CCG tests, after initiation, the cracks grew constantly and intergranularly before they accelerated to approach rupture. The creep crack growth rate is well described by C*. The relations between reference stress, failure time and steady crack growth rate are presented for the CCG tests. It is demonstrated that the failure stress due to CCG is considerably lower than the yield stress at 360 deg C. Consequently, the CCG will control the static strength of a reactor vessel. (orig.) 17 refs.

  6. IAEA international studies on irradiation embrittlement of reactor pressure vessel steels

    Energy Technology Data Exchange (ETDEWEB)

    Brumovsky, M. [Nuclear Research Institute Rez plc (Czech Republic); Steele, L.E. [Chief Scientific Investigator of the Programme, Springfield, VA (United States)

    1997-02-01

    In last 25 years, three phases a Co-operative Research Programme on Irradiation Embrittlement of Reactor Pressure Vessel Steels has been organized by the International Atomic Energy Agency. This programme started with eight countries in 1971 and finally 16 countries took part in phase III of the Programme in 1983. Several main efforts were put into preparation of the programme, but the principal task was concentrated on an international comparison of radiation damage characterization by different laboratories for steels of {open_quotes}old{close_quotes} (with high impurity contents) and {open_quotes}advanced{close_quotes} (with low impurity contents) types as well as on development of small scale fracture mechanics procedures applicable to reactor pressure vessel surveillance programmes. This year, a new programme has been opened, concentrated mostly on small scale fracture mechanics testing.

  7. RADIATION DOSIMETRY OF THE PRESSURE VESSEL INTERNALS OF THE HIGH FLUX BEAM REACTOR.

    Energy Technology Data Exchange (ETDEWEB)

    HOLDEN,N.E.; RECINIELLO,R.N.; HU,J.P.; RORER,D.C.

    2002-08-18

    In preparation for the eventual decommissioning of the High Flux Beam Reactor after the permanent removal of its fuel elements from the Brookhaven National Laboratory, both measurements and calculations of the decay gamma-ray dose rate have been performed for the reactor pressure vessel and vessel internal structures which included the upper and lower thermal shields, the transition plate, and the control rod blades. The measurements were made using Red Perspex{trademark} polymethyl methacrylate high-level film dosimeters, a Radcal ''peanut'' ion chamber, and Eberline's high-range ion chamber. To compare with measured gamma-ray dose rate, the Monte Carlo MCNP code and geometric progressive Microshield code were used to model the gamma transport and dose buildup.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-09-01

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

  9. Investigation of a Coolant Mixing Phenomena within the Reactor Pressure Vessel of a VVER-1000 Reactor with Different Simulation Tools

    Directory of Open Access Journals (Sweden)

    V. Sánchez

    2010-01-01

    Full Text Available The Institute of Neutron Physics and Reactor Technology (INR is involved in the qualification of coupled codes for reactor safety evaluations, aiming to improve their prediction capability and acceptability. In the frame of the VVER-1000 Coolant Transient Benchmark Phase 1, RELAP5/PARCS has been extensively assessed. Phase 2 of this benchmark was focused on both multidimensional thermal hydraulic phenomena and core physics. Plant data will be used to qualify the 3D models of TRACE and RELAP5/CFX, which were coupled for this purpose. The developed multidimensional models of the VVER-1000 reactor pressure vessel (RPV as well as the performed calculations will be described in detail. The predicted results are in good agreement with experimental data. It was demonstrated that the chosen 3D nodalization of the RPV is adequate for the description of the coolant mixing phenomena in a VVER-1000 reactor. Even though only a 3D coarse nodalization is used in TRACE, the integral results are comparable to those obtained by RELAP5/CFX.

  10. Roadmap for Nondestructive Evaluation of Reactor Pressure Vessel Research and Development by the Light Water Reactor Sustainability Program

    Energy Technology Data Exchange (ETDEWEB)

    Smith, Cyrus M [ORNL; Nanstad, Randy K [ORNL; Clayton, Dwight A [ORNL; Matlack, Katie [Georgia Institute of Technology; Ramuhalli, Pradeep [Pacific Northwest National Laboratory (PNNL); Light, Glenn [Southwest Research Institute, San Antonio

    2012-09-01

    The Department of Energy s (DOE) Light Water Reactor Sustainability (LWRS) Program is a five year effort which works to develop the fundamental scientific basis to understand, predict, and measure changes in materials and systems, structure, and components as they age in environments associated with continued long-term operations of existing commercial nuclear power reactors. This year, the Materials Aging and Degradation (MAaD) Pathway of this program has placed emphasis on emerging Non-Destructive Evaluation (NDE) methods which support these objectives. DOE funded Research and Development (R&D) on emerging NDE techniques to support commercial nuclear reactor sustainability is expected to begin next year. This summer, the MAaD Pathway invited subject matter experts to participate in a series of workshops which developed the basis for the research plan of these DOE R&D NDE activities. This document presents the results of one of these workshops which are the DOE LWRS NDE R&D Roadmap for Reactor Pressure Vessels (RPV). These workshops made a substantial effort to coordinate the DOE NDE R&D with that already underway or planned by the Electric Power Research Institute (EPRI) and the Nuclear Regulatory Commission (NRC) through their representation at these workshops.

  11. An experimental study of hypervapotron structure in external reactor vessel cooling

    Energy Technology Data Exchange (ETDEWEB)

    Zhao, Yufeng; Zhang, Ming [State Nuclear Power Technology R& D Center (Beijing), Beijing (China); Hou, Fangxin [Institute of Nuclear and New Energy Technology, Tsinghua University, Beijing (China); Gao, Tianfang [State Nuclear Power Technology R& D Center (Beijing), Beijing (China); Chen, Peipei, E-mail: chenpeipei@snptc.com.cn [State Power Investment Group Corporation, Beijing (China)

    2016-07-15

    Highlights: • Experiments are performed to study the application of hypervapotron in ERVC design. • CHF experiments on two surfaces are conducted under different flow conditions. • Hypervapotron improves CHF performance by 40–60% compared with smooth surface. • Visualization shows fin structure removes vapor mushroom for better liquid supply. - Abstract: In vessel retention (IVR) is one of the key strategies for many advanced LWR designs to mitigate postulated severe accidents. The success of IVR substantially relies on external reactor vessel cooling (ERVC) by which the decay heat is removed from the melt core in the reactor vessel lower head. The main challenge of IVR is to provide an adequate safety margin of ERVC against critical heat flux (CHF) of subcooled flow boiling in the reactor lower head flow channel. Due to uncertainties in corium melt pool configuration, large CHF margin of ERVC is usually required by regulatory authorities to demonstrate reliability of severe accident mitigation methods. Various CHF enhancement designs have been proposed and studied in literature. In this paper, an experimental study of hypervapotron structure as a novel design to improve CHF performance of ERVC is conducted. Hypervapotron is chosen as one of the potential engineering options for International Thermonuclear Experimental Reactor (ITER) program as a divertor structure to remove highly intense heat from fusion chamber. This study is to conduct CHF experiments at typical PWR ERVC working conditions. The CHF experiments are performed in a 30 mm by 61 mm rectangular flow channel with a 200 mm long heated surface along the flow direction. Both smooth and hypervapotron surface are tested at various inclination angles of the test section to simulate various positions of the reactor lower head. The hypervapotron is found to have a 40–60% CHF improvement compared with the smooth surface. The high speed visualization indicates that hypervapotron is able to

  12. Feasibility for development of a nuclear reactor pressure vessel flaw distribution: Sensitivity analyses and NDE (nondestructive evaluation) capability

    Energy Technology Data Exchange (ETDEWEB)

    Rosinski, S.T. (Sandia National Labs., Albuquerque, NM (USA)); Kennedy, E.L.; Foulds, J.R. (Failure Analysis Associates, Inc., Menlo Park, CA (USA))

    1990-01-01

    Pressurized water reactor pressure vessels operate under US Nuclear Regulatory Commission (NRC) rules and regulatory guides that are intended to maintain a low probability of vessel failure. The NRC has also addressed neutron embrittlement of pressurized water reactor pressure vessels by imposing regulations on plant operation. Plants failing to meet the operating criteria specified by these rules and regulations are required, among other things, to analytically demonstrate fitness for service in order to continue safe operation. The initial flaw size or distribution of initial vessel flaws is a key input to the required vessel integrity analyses. A fracture mechanics sensitivity study was performed to quantify the effect of the assumed flaw distribution on the predicted vessel performance under a specified pressurized thermal shock transient and to determine the critical crack size. Results of the analysis indicate that vessel performance in terms of the estimated probability of failure is very sensitive to the assumed flaw distribution. 20 refs., 3 figs., 2 tabs.

  13. The 5th surveillance testing for Kori unit 1 reactor vessel materials

    Energy Technology Data Exchange (ETDEWEB)

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

    2000-08-01

    Surveillance testing for reactor vessel materials is performed in order to evaluate the irradiation embrittlement due to neutrons during operation and set the condition of safe operation of nuclear reactor. The 5th surveillance testing was performed primarily by Korea Atomic Energy Research Institute and Westinhouse corporation partially involved in testing and calculation data evaluation in order to obtain reliable test result. Fast neutron fluences for capsule V, T, S, R and P were 5.087E+18, 1.115E+19, 1.228E+19, 2.988E+19, and 3.938E+19n/cm2, respectively. The bias factor, the ratio of calculation/measurement, was 0.940 for the 1st through 5th testing and the calculational uncertainty, 7% satisfied the requirement of USNRC Reg.Guide DG-1053, 20%. The best estimated neutron fluence for reactor vessel inside surface was 1.9846E+19n/cm{sup 2} based on the end of 17th fuel cycle and it was predicted that the fluences of vessel inside surface at 24, 32, 40 and 48EFPY would reach 3.0593E+19, 4.0695E+19, 5.0797E+19 and 6.0900E+19n/cm{sup 2} based on the current calculation. PTS analysis for Kori unit 1 showed that 27.93EFPY was the threshold value for 300 deg F requirement. 71 refs., 33 figs., 52 tabs. (Author)

  14. Fabrication Flaw Density and Distribution In Repairs to Reactor Pressure Vessel and Piping Welds

    Energy Technology Data Exchange (ETDEWEB)

    GJ Schuster, FA Simonen, SR Doctor

    2008-04-01

    The Pacific Northwest National Laboratory is developing a generalized fabrication flaw distribution for the population of nuclear reactor pressure vessels and for piping welds in U.S. operating reactors. The purpose of the generalized flaw distribution is to predict component-specific flaw densities. The estimates of fabrication flaws are intended for use in fracture mechanics structural integrity assessments. Structural integrity assessments, such as estimating the frequency of loss-of-coolant accidents, are performed by computer codes that require, as input, accurate estimates of flaw densities. Welds from four different reactor pressure vessels and a collection of archived pipes have been studied to develop empirical estimates of fabrication flaw densities. This report describes the fabrication flaw distribution and characterization in the repair weld metal of vessels and piping. This work indicates that large flaws occur in these repairs. These results show that repair flaws are complex in composition and sometimes include cracks on the ends of the repair cavities. Parametric analysis using an exponential fit is performed on the data. The relevance of construction records is established for describing fabrication processes and product forms. An analysis of these records shows there was a significant change in repair frequency over the years when these components were fabricated. A description of repair flaw morphology is provided with a discussion of fracture mechanics significance. Fabrication flaws in repairs are characterized using optimized-access, high-sensitivity nondestructive ultrasonic testing. Flaw characterizations are then validated by other nondestructive evaluation techniques and complemented by destructive testing.

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

    Energy Technology Data Exchange (ETDEWEB)

    Grounes, M.

    1966-03-15

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

  16. Isothermal and thermal–mechanical fatigue of VVER-440 reactor pressure vessel steels

    Energy Technology Data Exchange (ETDEWEB)

    Fekete, Balazs, E-mail: fekete.mm.bme@gmail.com [College of Dunaujvaros, Tancsics 1A, Dunaujvaros H-2400 (Hungary); Department of Applied Mechanics, Budapest University of Technology and Economics, Muegyetem 5, Budapest H-1111 (Hungary); Trampus, Peter [College of Dunaujvaros, Tancsics 1A, Dunaujvaros H-2400 (Hungary)

    2015-09-15

    Highlights: • We aimed to determine the thermomechanical behaviour of VVER reactor steels. • Material tests were developed and performed on GLEEBLE 3800 physical simulator. • Coffin–Manson curves and parameters were derived. • High accuracy of the strain energy based evaluation was found. • The observed dislocation evolution correlates with the mechanical behaviour. - Abstract: The fatigue life of the structural materials 15Ch2MFA (CrMoV-alloyed ferritic steel) and 08Ch18N10T (CrNi-alloyed austenitic steel) of VVER-440 reactor pressure vessel under completely reserved total strain controlled low cycle fatigue tests were investigated. An advanced test facility was developed for GLEEBLE-3800 physical simulator which was able to perform thermomechanical fatigue experiments under in-service conditions of VVER nuclear reactors. The low cycle fatigue results were evaluated with the plastic strain based Coffin–Manson law, and plastic strain energy based model as well. It was shown that both methods are able to predict the fatigue life of reactor pressure vessel steels accurately. Interrupted fatigue tests were also carried out to investigate the kinetic of the fatigue evolution of the materials. On these samples microstructural evaluation by TEM was performed. The investigated low cycle fatigue behavior can provide reference for remaining life assessment and lifetime extension analysis.

  17. A Visual Inspection System Development for the Reactor Vessel Bottom-mounted Instrument Penetrations

    Energy Technology Data Exchange (ETDEWEB)

    Jeong, Kyung Min; Choi, Young Su; Lee, Sung Uk; Choi, Chang Whan; Seo, Yong Chil; Kim, Chang Hoi; Jeon, Poong Woo; Kim, Seung Ho [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2005-07-01

    In April 2003, South Texas Project Unit 1 made a surprising discovery of boron acid leakage from two nozzles from a bare-metal examination of the reactor vessel bottom-mounted instrument penetrations during a routine refueling outage. A small powdery substance about 150mg was found on the outside of two instrument guide penetration nozzles on the bottom of the reactor. The primary coolant water of pressurized water reactors has caused cracking in penetrations with Alloy 600 through a process called primary water stress corrosion cracking. In South Korea, it is scheduled to conduct visual inspection of the outside of instrument guide penetration nozzles on the bottom of PWRs to confirm the integrity of reactor vessel. Inside the bottom head of KSNP, 45 penetration nozzles with the radius of about 7.5cm are mounted. The height of the nozzle located on the center is about 6 and the minimum distance between the nozzles is about 20 cm. In order to inspect the possible leakage of boron acid, two mobile robots are developed and described in this paper.

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

    Energy Technology Data Exchange (ETDEWEB)

    Kusunoki, Tsuyoshi; Odano, Naoteru; Yoritsune, Tsutomu; Ishida, Toshihisa [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment; Takahashi, Teruo [Energis, Co., Kobe, Hyogo (Japan); Nishimura, Hajime [Japan Marine Science and Technology Center, Yokosuka, Kanagawa (Japan); Tokunaga, Sango [Japan Deep Sea Technology Association, Tokyo (Japan)

    2001-07-01

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

  19. Issues of intergranular embrittlement of VVER-type nuclear reactors pressure vessel materials

    Science.gov (United States)

    Zabusov, O.

    2016-04-01

    In light of worldwide tendency to extension of service life of operating nuclear power plants - VVER-type in the first place - recently a special attention is concentrated on phenomena taking place in reactor pressure vessel materials that are able to lead to increased level of mechanical characteristics degradation (resistibility to brittle fracture) during long term of operation. Formerly the hardening mechanism of degradation (increase in the yield strength under influence of irradiation) mainly had been taken into consideration to assess pressure vessel service life limitations, but when extending the service life up to 60 years and more the non-hardening mechanism (intergranular embrittlement of the steels) must be taken into account as well. In this connection NRC “Kurchatov Institute” has initiated a number of works on investigations of this mechanism contribution to the total embrittlement of reactor pressure vessel steels. The main results of these investigations are described in this article. Results of grain boundary phosphorus concentration measurements in specimens made of first generation of VVER-type pressure vessels materials as well as VVER-1000 surveillance specimens are presented. An assessment of non-hardening mechanism contribution to the total ductile-to- brittle transition temperature shift is given.

  20. CFD analysis of moderator flow and temperature fields inside a vertical calandria vessel of nuclear reactor

    Energy Technology Data Exchange (ETDEWEB)

    Kansal, Anuj Kumar, E-mail: akansal@barc.gov.in [Bhabha Atomic Research Centre, Trombay, Mumbai 400085 (India); Joshi, Jyeshtharaj B., E-mail: jbjoshi@gmail.com [Homi Bhabha National Institute, Anushaktinagar, Mumbai 400094 (India); Maheshwari, Naresh Kumar, E-mail: nmahesh@barc.gov.in [Bhabha Atomic Research Centre, Trombay, Mumbai 400085 (India); Vijayan, Pallippattu Krishnan, E-mail: vijayanp@barc.gov.in [Bhabha Atomic Research Centre, Trombay, Mumbai 400085 (India)

    2015-06-15

    Highlights: • 3D CFD of vertical calandria vessel. • Spatial distribution of volumetric heat generation. • Effect of Archimedes number. • Non-dimensional analysis. - Abstract: Three dimensional computational fluid dynamics (CFD) analysis has been performed for the moderator flow and temperature fields inside a vertical calandria vessel of nuclear reactor under normal operating condition using OpenFOAM CFD code. OpenFOAM is validated by comparing the predicted results with the experimental data available in literature. CFD model includes the calandria vessel, calandria tubes, inlet header and outlet header. Analysis has been performed for the cases of uniform and spatial distribution of volumetric heat generation. Studies show that the maximum temperature in moderator is lower in the case of spatial distribution of heat generation as compared to that in the uniform heat generation in calandria. In addition, the effect of Archimedes number on maximum and average moderator temperature was investigated.

  1. The inclusion of weld residual stress in fracture margin assessments of embrittled nuclear reactor pressure vessels

    Energy Technology Data Exchange (ETDEWEB)

    Dickson, T.L.; Bass, B.R.; McAfee, W.J.

    1998-01-01

    Analyses were performed to determine the impact of weld residual stresses in a reactor pressure vessel (RPV) on (1) the generation of pressure temperature (P-T) curves required for maintaining specified fracture prevention margins during nuclear plant startup and shutdown, and (2) the conditional probability of vessel failure due to pressurized thermal shock (PTS) loading. The through wall residual stress distribution in an axially oriented weld was derived using measurements taken from a shell segment of a canceled RPV and finite element thermal stress analyses. The P-T curve derived from the best estimate load analysis and a t / 8 deep flaw, based on K{sub Ic}, was less limiting than the one derived from the current methodology prescribed in the ASME Boiler and Pressure Vessel Code. The inclusion of the weld residual stresses increased the conditional probability of cleavage fracture due to PTS loading by a factor ranging from 2 to 4.

  2. Proving test on the seismic reliability of nuclear power plant: PWR reactor containment vessel

    Energy Technology Data Exchange (ETDEWEB)

    Akiyama, Hiroshi; Yoshikawa, Teiichi; Ohno, Tokue; Yoshikawa, Eiji.

    1989-01-01

    Seismic reliability proving tests of nuclear power plant facilities are carried out by the Nuclear Power Engineering Test Center, using the large-scale, high-performance vibration table of Tadotsu Engineering Laboratory, and sponsored by the Ministry of International Trade and Industry. In 1982, the seismic reliability proving test of a PWR containment vessel was conducted using a test component of reduced scale 1/3.7. As a result of this test, the test component proved to have structural soundness against earthquakes, and at the same time its stable function was proved by leak tests which were carried out before and after the vibration test. In 1983, the detailed analysis and evaluation of these test results were carried out, and the analysis methods for evaluating strength against earthquakes were established. The seismic analysis and evaluation on the actual containment vessel were then performed using these analysis methods, and the safety and reliability of the PWR reactor containment vessel were confirmed.

  3. Development of seismic sloshing analysis method of liquid coolant sodium in the KALIMER reactor vessel including several cylindrical components

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Jae Han; Yoo, Bong

    2000-11-01

    It is important to establish a highly accurate technique of evaluating the sloshing behavior of liquid sodium coolant during earthquake for structural integrity of KALIMER reactor vessel and internals. The analysis procedure of sloshing behaviors is established using finite element computer program ANSYS, and the effectiveness of the procedure is confirmed by comparison with theoretical and experimental results in the literature. The analysis results agree well with experimental ones. Based on the procedure, the sloshing characteristics of liquid sodium coolant in the KALIMER reactor vessel including reactor internal components are evaluated. The maximum response height of sodium free surface at the reactor vessel is about 55cm when subjected to horizontal safe shutdown earthquake (SSE) of 0.3g for seismically isolated reactor building.

  4. Helium Leak Detection of Vessels in Fuel Transfer Cell (FTC) of Prototype Fast Breeder Reactor (PFBR)

    Science.gov (United States)

    Dutta, N. G.

    2012-11-01

    Bharatiya Nabhikiya Vidyut Nigam (BHAVINI) is engaged in construction of 500MW Prototype Fast Breeder Reactor (PFBR) at Kalpak am, Chennai. In this very important and prestigious national programme Special Product Division (SPD) of M/s Kay Bouvet Engg.pvt. ltd. (M/s KBEPL) Satara is contributing in a major way by supplying many important sub-assemblies like- Under Water trolley (UWT), Airlocks (PAL, EAL) Container and Storage Rack (CSR) Vessels in Fuel Transfer Cell (FTC) etc for PFBR. SPD of KBEPL caters to the requirements of Government departments like - Department of Atomic Energy (DAE), BARC, Defense, and Government undertakings like NPCIL, BHAVINI, BHEL etc. and other precision Heavy Engg. Industries. SPD is equipped with large size Horizontal Boring Machines, Vertical Boring Machines, Planno milling, Vertical Turret Lathe (VTL) & Radial drilling Machine, different types of welding machines etc. PFBR is 500 MWE sodium cooled pool type reactor in which energy is produced by fissions of mixed oxides of Uranium and Plutonium pellets by fast neutrons and it also breeds uranium by conversion of thorium, put along with fuel rod in the reactor. In the long run, the breeder reactor produces more fuel then it consumes. India has taken the lead to go ahead with Fast Breeder Reactor Programme to produce electricity primarily because India has large reserve of Thorium. To use Thorium as further fuel in future, thorium has to be converted in Uranium by PFBR Technology.

  5. Thick SS316 materials TIG welding development activities towards advanced fusion reactor vacuum vessel applications

    Science.gov (United States)

    Kumar, B. Ramesh; Gangradey, R.

    2012-11-01

    Advanced fusion reactors like ITER and up coming Indian DEMO devices are having challenges in terms of their materials design and fabrication procedures. The operation of these devices is having various loads like structural, thermo-mechanical and neutron irradiation effects on major systems like vacuum vessel, divertor, magnets and blanket modules. The concept of double wall vacuum vessel (VV) is proposed in view of protecting of major reactor subsystems like super conducting magnets, diagnostic systems and other critical components from high energy 14 MeV neutrons generated from fusion plasma produced by D-T reactions. The double walled vacuum vessel is used in combination with pressurized water circulation and some special grade borated steel blocks to shield these high energy neutrons effectively. The fabrication of sub components in VV are mainly used with high thickness SS materials in range of 20 mm- 60 mm of various grades based on the required protocols. The structural components of double wall vacuum vessel uses various parts like shields, ribs, shells and diagnostic vacuum ports. These components are to be developed with various welding techniques like TIG welding, Narrow gap TIG welding, Laser welding, Hybrid TIG laser welding, Electron beam welding based on requirement. In the present paper the samples of 20 mm and 40 mm thick SS 316 materials are developed with TIG welding process and their mechanical properties characterization with Tensile, Bend tests and Impact tests are carried out. In addition Vickers hardness tests and microstructural properties of Base metal, Heat Affected Zone (HAZ) and Weld Zone are done. TIG welding application with high thick SS materials in connection with vacuum vessel requirements and involved criticalities towards welding process are highlighted.

  6. Isothermal and thermal-mechanical fatigue of VVER-440 reactor pressure vessel steels

    Science.gov (United States)

    Fekete, Balazs; Trampus, Peter

    2015-09-01

    The fatigue life of the structural materials 15Ch2MFA (CrMoV-alloyed ferritic steel) and 08Ch18N10T (CrNi-alloyed austenitic steel) of VVER-440 reactor pressure vessel under completely reserved total strain controlled low cycle fatigue tests were investigated. An advanced test facility was developed for GLEEBLE-3800 physical simulator which was able to perform thermomechanical fatigue experiments under in-service conditions of VVER nuclear reactors. The low cycle fatigue results were evaluated with the plastic strain based Coffin-Manson law, and plastic strain energy based model as well. It was shown that both methods are able to predict the fatigue life of reactor pressure vessel steels accurately. Interrupted fatigue tests were also carried out to investigate the kinetic of the fatigue evolution of the materials. On these samples microstructural evaluation by TEM was performed. The investigated low cycle fatigue behavior can provide reference for remaining life assessment and lifetime extension analysis.

  7. Effects of irradiation at lower temperature on the microstructure of Cr-Mo-V-alloyed reactor pressure vessel steel

    Energy Technology Data Exchange (ETDEWEB)

    Grosse, M.; Boehmert, J.; Gilles, R. [Hahn-Meitner-Institut Berlin GmbH (Germany)

    1998-10-01

    The microstructural damage process due to neutron irradiation [1] proceeds in two stages: - formation of displacement cascades - evolution of the microstructure by defect reactions. Continuing our systematic investigation about the microstructural changes of Russian reactor pressure vessel steel due to neutron irradiation the microstructure of two laboratory heats of the VVER 440-type reactor pressure vessel steel after irradiation at 60 C was studied by small angle neutron scattering (SANS). 60 C-irradiation differently changes the irradiation-induced microstructure in comparison with irradiation at reactor operation temperature and can, thus, provide new insights into the mechanisms of the irradiation damage. (orig.)

  8. Standard Test Method for Application and Analysis of Helium Accumulation Fluence Monitors for Reactor Vessel Surveillance, E706 (IIIC)

    CERN Document Server

    American Society for Testing and Materials. Philadelphia

    2007-01-01

    1.1 This test method describes the concept and use of helium accumulation for neutron fluence dosimetry for reactor vessel surveillance. Although this test method is directed toward applications in vessel surveillance, the concepts and techniques are equally applicable to the general field of neutron dosimetry. The various applications of this test method for reactor vessel surveillance are as follows: 1.1.1 Helium accumulation fluence monitor (HAFM) capsules, 1.1.2 Unencapsulated, or cadmium or gadolinium covered, radiometric monitors (RM) and HAFM wires for helium analysis, 1.1.3 Charpy test block samples for helium accumulation, and 1.1.4 Reactor vessel (RV) wall samples for helium accumulation. This standard does not purport to address all of the safety concerns, if any, associated with its use. It is the responsibility of the user of this standard to establish appropriate safety and health practices and determine the applicability of regulatory limitations prior to use.

  9. Review of pressurized thermal shock studies of large scale reactor pressure vessels in Hungary

    Directory of Open Access Journals (Sweden)

    Tamás Fekete

    2016-03-01

    Full Text Available In Hungary, four nuclear power units were constructed more than 30 years ago; they are operating to this day. In every unit, VVER-440 V213-type light-water cooled, light-water moderated, ressurized water reactors are in operation. Since the mid-1980s, numerous researches in the field of Pressurized Thermal Shock (PTS analyses of Reactor Pressure Vessels (RPVs have been conducted in Hungary; in all of them, the concept of structural integrity was the basis of research and development. During this time, four large PTS studies with industrial relevance have been completed in Hungary. Each used different objectives and guides, and the analysis methodology was also changing. This paper gives a comparative review of the methodologies used in these large PTS Structural Integrity Analysis projects, presenting the latest results as well

  10. Thermal-hydraulic characteristics in a tokamak vacuum vessel of fusion reactor after transient events occurred

    Energy Technology Data Exchange (ETDEWEB)

    Takase, Kazuyuki; Kunugi, Tomoaki [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment; Seki, Yasushi

    1997-12-31

    The thermal-hydraulic characteristics in a vacuum vessel (VV) of fusion reactor under the ingress-of-coolant-event (ICE) or loss-of-vacuum-event (LOVA) condition were carried out to investigate experimentally the thermofluid safety in the International Thermonuclear Experimental Reactor (ITER) under transient events. In the ICE experiments, the pressure rise and wall temperatures in the VV were measured and the performance of a suppression tank was confirmed. In the LOVA experiments, the exchange time inside the VV from the vacuum to be the atmospheric pressure was measured for various breach size and the exchange flow rates through the breaches of the VV under the atmospheric pressure conditions were clarified. (author)

  11. Positron annihilation study of Fe-ion irradiated reactor pressure vessel model alloys

    Science.gov (United States)

    Chen, L.; Li, Z. C.; Schut, H.; Sekimura, N.

    2016-01-01

    The degradation of reactor pressure vessel steels under irradiation, which results from the hardening and embrittlement caused by a high number density of nanometer scale damage, is of increasingly crucial concern for safe nuclear power plant operation and possible reactor lifetime prolongation. In this paper, the radiation damage in model alloys with increasing chemical complexity (Fe, Fe-Cu, Fe-Cu-Si, Fe-Cu-Ni and Fe-Cu-Ni-Mn) has been studied by Positron Annihilation Doppler Broadening spectroscopy after 1.5 MeV Fe-ion implantation at room temperature or high temperature (290 oC). It is found that the room temperature irradiation generally leads to the formation of vacancy-type defects in the Fe matrix. The high temperature irradiation exhibits an additional annealing effect for the radiation damage. Besides the Cu-rich clusters observed by the positron probe, the results show formation of vacancy-Mn complexes for implantation at low temperatures.

  12. Application of the small punch test to reactor pressure vessel integrity

    Energy Technology Data Exchange (ETDEWEB)

    Rosinski, S.T. [EPRI Nondestructive Evaluation Center, Charlotte, NC (United States); Viswanathan, R. [EPRI, Palo Alto, CA (United States); Foulds, J.R. [Failure Analysis Associates, Inc., Menlo Park, CA (United States)

    1998-07-01

    Based on prior success with fossil plant steels, EPRI is investigating the feasibility of applying the Small Punch test to determine the fracture toughness (K{sub ic}) of irradiated reactor pressure vessel (RPV) materials. The small punch test specimen is sufficiently small to alleviate future surveillance material availability concerns, as well as provide a means of direct vessel material interrogation by non-disruptive miniature sample removal and testing. A limited series of small punch tests on unirradiated and irradiated RPV steel materials has shown that the method can be used to estimate ductile-to-brittle transition temperatures and to determine the material fracture toughness (K{sub lc}, J{sub lc}). The results to date are described and the experimental difficulties that need to be resolved in achieving valid results are identified. (authors)

  13. Safety assessment of in-vessel vapor explosion loads in next generation reactor

    Energy Technology Data Exchange (ETDEWEB)

    Bang, Kwang Hyun; Cho, Jong Rae; Choi, Byung Uk; Kim, Ki Yong; Lee, Kyung Jung [Korea Maritime University, Busan (Korea); Park, Ik Kyu [Seoul National University, Seoul (Korea)

    1998-12-01

    A safety assessment of the reactor vessel lower head integrity under in-vessel vapor explosion loads has been performed. The premixing and explosion calculations were performed using TRACER-II code. Using the calculated explosion pressures imposed on the lower head inner wall, strain calculations were performed using ANSYS code. The explosion analyses show that the explosion impulses are not altered significantly by the uncertain parameters of triggering location and time, fuel and vapor volume fractions in uniform premixture bounding calculations within the conservative ranges. Strain analyses using the calculated pressure loads on the lower head inner wall show that the vapor explosion-induced lower head failure is physically unreasonable. The static analysis using the conservative explosion-end pressure of 7,246 psia shows that the maximum equivalent strain is 4.3% at the bottom of lower head, which is less than the allowable threshold value of 11%. (author). 24 refs., 40 figs., 3 tabs.

  14. Preliminary materials selection issues for the next generation nuclear plant reactor pressure vessel.

    Energy Technology Data Exchange (ETDEWEB)

    Natesan, K.; Majumdar, S.; Shankar, P. S.; Shah, V. N.; Nuclear Engineering Division

    2007-03-21

    In the coming decades, the United States and the entire world will need energy supplies to meet the growing demands due to population increase and increase in consumption due to global industrialization. One of the reactor system concepts, the Very High Temperature Reactor (VHTR), with helium as the coolant, has been identified as uniquely suited for producing hydrogen without consumption of fossil fuels or the emission of greenhouse gases [Generation IV 2002]. The U.S. Department of Energy (DOE) has selected this system for the Next Generation Nuclear Plant (NGNP) Project, to demonstrate emissions-free nuclear-assisted electricity and hydrogen production within the next 15 years. The NGNP reference concepts are helium-cooled, graphite-moderated, thermal neutron spectrum reactors with a design goal outlet helium temperature of {approx}1000 C [MacDonald et al. 2004]. The reactor core could be either a prismatic graphite block type core or a pebble bed core. The use of molten salt coolant, especially for the transfer of heat to hydrogen production, is also being considered. The NGNP is expected to produce both electricity and hydrogen. The process heat for hydrogen production will be transferred to the hydrogen plant through an intermediate heat exchanger (IHX). The basic technology for the NGNP has been established in the former high temperature gas reactor (HTGR) and demonstration plants (DRAGON, Peach Bottom, AVR, Fort St. Vrain, and THTR). In addition, the technologies for the NGNP are being advanced in the Gas Turbine-Modular Helium Reactor (GT-MHR) project, and the South African state utility ESKOM-sponsored project to develop the Pebble Bed Modular Reactor (PBMR). Furthermore, the Japanese HTTR and Chinese HTR-10 test reactors are demonstrating the feasibility of some of the planned components and materials. The proposed high operating temperatures in the VHTR place significant constraints on the choice of material selected for the reactor pressure vessel for

  15. Robinson 2 reactor vessel: pressurized thermal shock analysis for a small-break LOCA

    Energy Technology Data Exchange (ETDEWEB)

    Marston, T.; Griesbach, T.; Chao, J.; Chexal, B.; Norris, D.; Nickell, B.; Layman, B.

    1984-08-01

    A best-estimate Pressurized Thermal Shock (PTS) analysis was performed for a three-inch diameter hot-leg small-break loss-of-coolant accident for the Robinson 2 plant. This plant specific analysis was performed using EPRI's linked set of codes for PTS analysis. The analysis shows that with the H.B. Robinson 2 reactor pressure vessel, a hot-leg small-break loss-of-coolant accident does not pose a significant health or safety concern to the public for at least 40 years of operation.

  16. Improvement of remote control system of automatic ultrasonic equipment for inspection of reactor pressure vessel

    Energy Technology Data Exchange (ETDEWEB)

    Cheong, Yong Moo; Jung, H. K.; Joo, Y. S.; Koo, K. M.; Hyung, H.; Sim, C. M.; Gong, U. S.; Kim, S. H.; Lee, J. P.; Rhoo, H. C.; Kim, M. S.; Ryoo, S. K.; Choi, C. H.; Oh, K. I

    1999-12-01

    One of the important issues related to the nuclear safety is in-service inspection of reactor pressure vessel (RPV). A remote controlled automatic ultrasonic method is applied to the inspection. At present the automatic ultrasonic inspection system owned by KAERI is interrupted due to degradation of parts. In order to resume field inspection new remote control system for the equipment was designed and installed to the existing equipment. New ultrasonic sensors and their modules for RPV inspection were designed and fabricated in accordance with the new requirements of the inspection codes. Ultrasonic sensors were verified for the use in the RPV inspection. (autho0008.

  17. Distribution of properties in nuclear reactor vessel shells in the unirradiated state

    Science.gov (United States)

    Skundin, M. A.; Chernobaeva, A. A.; Zhurko, D. A.; Krasikov, E. A.; Medvedev, K. I.

    2013-04-01

    The distributions of the chemical composition, the strength characteristics, and critical ductile-brittle transition temperature T cr are studied in the axial, radial, and tangential directions of the material of a test ring cut from a standard forging used for a VVER-1000 reactor vessel shell. The values of T cr of specimens cut from the test ring are shown to be well below those of the internal volume of the shell, which can explain the substantial scatter of the results obtained on reference specimens cut from the base metal.

  18. Evaluation of dynamic fracture toughness for Yong Gwang unit 5 reactor pressure vessel materials (Baseline Tests)

    Energy Technology Data Exchange (ETDEWEB)

    Chi Se Hwan; Kim, Joo Hag; Hong, Jun Hwa; Kwon, Sun Chil; Lee, Bong Sang [Korea Atomic Energy Research Institute, Taejon (Korea)

    1999-10-01

    The dynamic fracture toughness (K{sub d}) of intermediate shell and its weld in SA 508 CI. 3 Yong Gwang 5 reactor pressure vessel was determined and evaluated. Precracked thirty six Charpy specimens were tested by using an instrumented impact tester. The purpose of present work is to evaluate and confirm the un-irradiated dynamic fracture toughness and to provide pre-irradiation baseline data for future evaluation on dynamic fracture toughness change during operation. 18 refs., 5 figs., 5 tabs. (Author)

  19. Cutting Technology for Decommissioning of the Reactor Pressure Vessels in Nuclear Power Plants

    Energy Technology Data Exchange (ETDEWEB)

    Jeong, Kwan Seong; Kim, Geun Ho; Moon, Jei Kwon; Choi, Byung Seon [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2012-05-15

    Lots of nuclear power plants have been decommissioned during the last 2 decades. An essential part of this work is the dismantling of the Reactor Pressure Vessel and its Internals. For this purpose a wide variety of different cutting technologies have been developed, adapted and applied. A detailed introduction to Plasma Arc cutting, Contact Arc Metal cutting and Abrasive Water Suspension Jet cutting is given, as it turned out that these cutting technologies are particularly suitable for these type of segmentation work. A comparison of these technologies including gaseous emissions, cutting power, manipulator requirements as well as selected design approaches are given. Process limits as well as actual limits of application are presented

  20. Next Generation Nuclear Plant Reactor Pressure Vessel Materials Research and Development Plan (PLN-2803)

    Energy Technology Data Exchange (ETDEWEB)

    J. K. Wright; R. N. Wright

    2010-07-01

    The U.S. Department of Energy (DOE) has selected the High-Temperature Gas-cooled Reactor (HTGR) design for the Next Generation Nuclear Plant (NGNP) Project. The NGNP will demonstrate the use of nuclear power for electricity and hydrogen production, with an outlet gas temperature in the range of 750°C, and a 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, Tri-Isotopic (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. 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 setting the stage for an economically viable deployment of the new technology in the commercial sector soon after 2020. This technology development plan details the additional research and development (R&D) required to design and license the NGNP RPV, assuming that A 508/A 533 is the material of construction. The majority of additional information that is required is related to long-term aging behavior at NGNP vessel temperatures, which are somewhat above those commonly encountered in the existing database from LWR experience. Additional data are also required for the anticipated NGNP environment. An assessment of required R&D for a Grade 91 vessel has been retained from the first revision of the R&D plan in Appendix B in somewhat less detail. Considerably more development is required for this steel compared to A 508/A 533 including additional irradiation testing for expected NGNP operating temperatures, high-temperature mechanical properties, and extensive studies of long-term microstructural stability.

  1. Next Generation Nuclear Plant Reactor Pressure Vessel Materials Research and Development Plan (PLN-2803)

    Energy Technology Data Exchange (ETDEWEB)

    J. K. Wright; R. N. Wright

    2008-04-01

    The U.S. 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, Tri-Isotopic-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 Program is responsible for performing research and development 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 setting the stage for an economically viable deployment of the new technology in the commercial sector soon after 2020. Studies of potential Reactor Pressure Vessel (RPV) steels have been carried out as part of the pre-conceptual design studies. These design studies generally focus on American Society of Mechanical Engineers (ASME) Code status of the steels, temperature limits, and allowable stresses. Three realistic candidate materials have been identified by this process: conventional light water reactor RPV steels A508/533, 2¼Cr-1Mo in the annealed condition, and modified 9Cr 1Mo ferritic martenistic steel. Based on superior strength and higher temperature limits, the modified 9Cr-1Mo steel has been identified by the majority of design engineers as the preferred choice for the RPV. All of the vendors have

  2. UT digitized data processing for in service inspection of pressurized water reactor vessels

    Energy Technology Data Exchange (ETDEWEB)

    Lasserre, F.; Hernandez, L. [Intercontrole, Rungis (France); Paradis, L. [CEA/CEREM, 91191, Gifs/Yvette cedex (France)

    1998-03-01

    Pressurized reactor vessels in France have been examined from the inside with ultrasonic focused transducers since the very first inspection (Gagnor and Levy (1993)). The developments carried out in collaboration with the French Atomic Energy Commission (CEA) to improve the characterization of flaws detected in the body of the vessels or in the nozzles, in the vicinity of the inner or the outer surfaces now have application throughout the CIVAMIS software. The processing modules of CIVAMIS, which are implemented on site since 1994 and used by INTERCONTROLE during the in service inspections of the French PWR vessels, allow full characterization of these specific flaws. The first module is devoted to the characterization of defects located near the outer surface of the vessel or the bottom head welds (OSD module). It includes the modeling software MEPHISTOMIS which predicts the echoes coming from the interaction between the ultrasonic beam and the defects. The second module of CIVAMIS (inner surface defect module called ISD), applied to the analysis of flaws expected near the inner surface of the vessels, has been used during performance demonstration exercises on qualification mock-ups, and also on-site in five expert appraisals since its qualification in 1995. The third module available on the system has beendeveloped and qualified for the analysis of flaws likely to appear near the inner surface of the no zzles. This module, named `undercladding crack defect` (UCD) module, provides the operators with a set of pre-defined processing configurations well adapted to the characteristics of the transducers. (orig.) 11 refs.

  3. Probabilistic Fracture Mechanics of Reactor Pressure Vessels with Populations of Flaws

    Energy Technology Data Exchange (ETDEWEB)

    Spencer, Benjamin [Idaho National Lab. (INL), Idaho Falls, ID (United States); Backman, Marie [Univ. of Tennessee, Knoxville, TN (United States); Williams, Paul [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Hoffman, William [Idaho National Lab. (INL), Idaho Falls, ID (United States); Alfonsi, Andrea [Idaho National Lab. (INL), Idaho Falls, ID (United States); Dickson, Terry [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Bass, B. Richard [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Klasky, Hilda [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2016-09-01

    This report documents recent progress in developing a tool that uses the Grizzly and RAVEN codes to perform probabilistic fracture mechanics analyses of reactor pressure vessels in light water reactor nuclear power plants. The Grizzly code is being developed with the goal of creating a general tool that can be applied to study a variety of degradation mechanisms in nuclear power plant components. Because of the central role of the reactor pressure vessel (RPV) in a nuclear power plant, particular emphasis is being placed on developing capabilities to model fracture in embrittled RPVs to aid in the process surrounding decision making relating to life extension of existing plants. A typical RPV contains a large population of pre-existing flaws introduced during the manufacturing process. The use of probabilistic techniques is necessary to assess the likelihood of crack initiation at one or more of these flaws during a transient event. This report documents development and initial testing of a capability to perform probabilistic fracture mechanics of large populations of flaws in RPVs using reduced order models to compute fracture parameters. The work documented here builds on prior efforts to perform probabilistic analyses of a single flaw with uncertain parameters, as well as earlier work to develop deterministic capabilities to model the thermo-mechanical response of the RPV under transient events, and compute fracture mechanics parameters at locations of pre-defined flaws. The capabilities developed as part of this work provide a foundation for future work, which will develop a platform that provides the flexibility needed to consider scenarios that cannot be addressed with the tools used in current practice.

  4. Sintered silicon carbide: a new ceramic vessel material for microwave chemistry in single-mode reactors.

    Science.gov (United States)

    Gutmann, Bernhard; Obermayer, David; Reichart, Benedikt; Prekodravac, Bojana; Irfan, Muhammad; Kremsner, Jennifer M; Kappe, C Oliver

    2010-10-25

    Silicon carbide (SiC) is a strongly microwave absorbing chemically inert ceramic material that can be utilized at extremely high temperatures due to its high melting point and very low thermal expansion coefficient. Microwave irradiation induces a flow of electrons in the semiconducting ceramic that heats the material very efficiently through resistance heating mechanisms. The use of SiC carbide reaction vessels in combination with a single-mode microwave reactor provides an almost complete shielding of the contents inside from the electromagnetic field. Therefore, such experiments do not involve electromagnetic field effects on the chemistry, since the semiconducting ceramic vial effectively prevents microwave irradiation from penetrating the reaction mixture. The involvement of electromagnetic field effects (specific/nonthermal microwave effects) on 21 selected chemical transformations was evaluated by comparing the results obtained in microwave-transparent Pyrex vials with experiments performed in SiC vials at the same reaction temperature. For most of the 21 reactions, the outcome in terms of conversion/purity/product yields using the two different vial types was virtually identical, indicating that the electromagnetic field had no direct influence on the reaction pathway. Due to the high chemical resistance of SiC, reactions involving corrosive reagents can be performed without degradation of the vessel material. Examples include high-temperature fluorine-chlorine exchange reactions using triethylamine trihydrofluoride, and the hydrolysis of nitriles with aqueous potassium hydroxide. The unique combination of high microwave absorptivity, thermal conductivity, and effusivity on the one hand, and excellent temperature, pressure and corrosion resistance on the other hand, makes this material ideal for the fabrication of reaction vessels for use in microwave reactors.

  5. Interactions between dislocations and irradiation-induced defects in light water reactor pressure vessel steels

    Science.gov (United States)

    Jumel, Stéphanie; Van Duysen, Jean-Claude; Ruste, Jacky; Domain, Christophe

    2005-11-01

    The REVE project (REactor for Virtual Experiments) is an international effort aimed at developing tools to simulate irradiation effects in light water reactors materials. In the framework of this project, a European team developed a first tool, called RPV-1 designed for reactor pressure vessel steels. This article is the third of a series dedicated to the presentation of the codes and models used to build RPV-1. It describes the simplified approach adopted to simulate the irradiation-induced hardening. This approach relies on a characterization of the interactions between a screw dislocation and irradiation-induced defects from molecular dynamics simulations. The pinning forces exerted by the defects on the dislocation were estimated from the obtained results and some hypotheses. In RPV-1, these forces are used as input parameters of a Foreman and Makin-type code, called DUPAIR, to simulate the irradiation-induced hardening at 20 °C. The relevance of the proposed approach was validated by the comparison with experimental results. However, this work has to be considered as an initial step to facilitate the development of a first tool to simulate irradiation effects. It can be improved by many ways (e.g. by use of dislocation dynamics code).

  6. Nanostructure evolution of neutron-irradiated reactor pressure vessel steels: Revised Object kinetic Monte Carlo model

    Science.gov (United States)

    Chiapetto, M.; Messina, L.; Becquart, C. S.; Olsson, P.; Malerba, L.

    2017-02-01

    This work presents a revised set of parameters to be used in an Object kinetic Monte Carlo model to simulate the microstructure evolution under neutron irradiation of reactor pressure vessel steels at the operational temperature of light water reactors (∼300 °C). Within a "grey-alloy" approach, a more physical description than in a previous work is used to translate the effect of Mn and Ni solute atoms on the defect cluster diffusivity reduction. The slowing down of self-interstitial clusters, due to the interaction between solutes and crowdions in Fe is now parameterized using binding energies from the latest DFT calculations and the solute concentration in the matrix from atom-probe experiments. The mobility of vacancy clusters in the presence of Mn and Ni solute atoms was also modified on the basis of recent DFT results, thereby removing some previous approximations. The same set of parameters was seen to predict the correct microstructure evolution for two different types of alloys, under very different irradiation conditions: an Fe-C-MnNi model alloy, neutron irradiated at a relatively high flux, and a high-Mn, high-Ni RPV steel from the Swedish Ringhals reactor surveillance program. In both cases, the predicted self-interstitial loop density matches the experimental solute cluster density, further corroborating the surmise that the MnNi-rich nanofeatures form by solute enrichment of immobilized small interstitial loops, which are invisible to the electron microscope.

  7. Reactor cover system improvements. Final report FY 1978. [Seal, bearing, support, deck, vessel support

    Energy Technology Data Exchange (ETDEWEB)

    McEdwards, J.A.; Matteras, J.L.

    1978-01-01

    The effort in the reactor cover study area resulted in design recommendations for the vessel support, the deck, and the bearing and seals. Sixteen configurations of bearings and seals were examined as part of this study. The selected concepts use a double inflatable seal plus a sodium dip seal. Six different deck configurations were considered as part of this study. The most attractive of these concepts is the conical deck. Five different vessel support concepts were considered. Of these, the U ring appears to be the most attractive. Significant findings are the following: (1) verified that passive cooling of the deck and support lead to acceptable temperatures; (2) the assembly tolerances can be loosened for lower fabrication cost and easier operation while meeting positional and sealing requirements; (3) determined that the conical deck is the most effective deck configuration; (4) determined that the U ring is the most effective vessel support configuration; (5) selected a bearing and seal approach that gives effective gas sealing, adequate control of sodium frost, and easy maintenance.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2001-03-01

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

  9. Analysis of fluid mixing characteristics in reactor vessel downcomer using Theofanous and Wallis` mixing model. (DVI Case)

    Energy Technology Data Exchange (ETDEWEB)

    Cho, Bong Hyun; Kim, Hwan Yeol; Kang, Hyung Seok; Bae, Yoon Young

    1997-05-01

    Direct injection of emergency core cooling water into the reactor vessel downcomer annulus (DVI) is an unique feature of the four-train safety injection system of Korean Next Generation Reactor(KNGR). In this study, in order to evaluate the fluid mixing characteristics of the injected water for DVI case, we have suggested for application to DVI, Theofanous` regional mixing model and Wallis` experiments of flow regimes for injection water to the annulus. Theofanous`model was developed as a fluid mixing model in reactor vessel downcomer for the case of Cold Leg Injection(CLI). We have established a procedure for calculating fluid mixing temperature, calculated the mixing temperature for SBLOCA and MSLB, and compared them to those of CLI. In general, the fluid temperatures across the reactor vessel beltline are higher than 110 deg F, the RT{sub NDT} of EOL for reactor vessel material, and the values are within the acceptable limits of PTS concern. (author). 6 tabs., 21 figs., 11 refs.

  10. Inspection of dissimilar metal welds in reactor pressure vessels in Spain

    Energy Technology Data Exchange (ETDEWEB)

    Gadea, J.R.; Regidor, J.J.; Pelaez, J.A.; Serrano, P. [Tecnatom, S.A., San Sebastian de los Reyes, Madrid (Spain)

    2011-07-01

    MRP-139 recommendations for inspection of dissimilar metal (DM) welds in PWR vessels were launched in the last years in the USA. Basically, it increases the frequency of the examinations in these type of welds, with major emphasis in the hot loops, adding one intermediate inspection at the ten years interval in outlet nozzles. The spanish nuclear power plants (NPP's) have begun the implementation of this type of inspections on the vessel nozzles DM welds. As this type of inspections could have an impact in the critical path duration of the outage, it is necessary the use of a mechanical equipment able to examine the nozzles DM welds in a short vessel occupation time (VOT) with high quality, qualified techniques and minimum requirements of the refuelling platform. Tecnatom undertook the design and development of a new more advanced equipment, named TENIS-DM, for implementing the reactor pressure vessel (RPV) nozzles examination. This equipment was designed in order to accomplish the stringent requirements and the updated examination techniques; it was used for the inspection of the DM welds of Asco 1 NPP inlet and outlet nozzles in March 2011. Examination techniques and procedures were qualified through the GRUVAL validation program, based on ENIC methodology. Mechanical scanner was equipped with a large number of examination probes, and TV cameras -for visual inspection and also for monitoring the ultrasonic inspections. A remote operated submarine was also used to give support to the operational personnel during the manipulation of the equipment and its movements from one nozzle to the others. During two months before the inspection, tests of the complete inspection system were made on a nozzle mock-up installed in a 4 meters deep well at Tecnatom's facilities; this scenario was also used during the training sessions of the inspection crew. The defined technical and practical objectives were achieved: use of qualified techniques and minimal impact on the

  11. Creep-Fatigue Damage Evaluation of a Model Reactor Vessel and Reactor Internals of Sodium Test Facility according to ASME-NH and RCC-MRx Codes

    Energy Technology Data Exchange (ETDEWEB)

    Lim, Dong-Won; Lee, Hyeong-Yeon; Eoh, Jae-Hyuk; Son, Seok-Kwon; Kim, Jong-Bum; Jeong, Ji-Young [KAERI, Daejeon (Korea, Republic of)

    2016-05-15

    The objective of the STELLA-2 is to support the specific design approval for PGSFR by synthetic reviews of key safety issues and code validations through the integral effect tests. Due to its high temperature operation in SFRs (and in a testing facility) up to 550 °C, thermally induced creep-fatigue damage is very likely in components including a reactor vessel, reactor internals (interior structures), heat exchangers, pipelines, etc. In this study, structural integrity of the components such as reactor vessel and internals in STELLA-2 has been evaluated against creep-fatigue failures at a concept-design step. As 2D analysis yields far conservative results, a realistic 3D simulation is performed by a commercial software. A design integrity guarding against a creep-fatigue damage failure operating at high temperature was evaluated for the reactor vessel with its internal structure of the STELLA-2. Both the high temperature design codes were used for the evaluation, and results were compared. All the results showed the vessel as a whole is safely designed at the given operating conditions, while the ASME-NH gives a conservative evaluation.

  12. CFD analysis of PWR core top and reactor vessel upper plenum internal subdomain models

    Energy Technology Data Exchange (ETDEWEB)

    Kao, Min-Tsung; Wu, Chung-Yun [National Tsing Hua University, Hsinchu 30043, Taiwan (China); Chieng, Ching-Chang, E-mail: cchieng@ess.nthu.edu.tw [National Tsing Hua University, Hsinchu 30043, Taiwan (China); Xu Yiban; Yuan Kun; Dzodzo, Milorad; Conner, Michael; Beltz, Steven; Ray, Sumit; Bissett, Teresa [Westinghouse Electric Company, Cranberry Township, PA 16066 (United States)

    2011-10-15

    Highlights: > The paper develops a CFD flow model for upper portion of AP1000 and determines how lateral flow in the top core and upper plenum. > Mesh sensitivities and geometrical modification strategies give the guidelines to reduce the size of overall computation mesh. > Pressure drop measurement data act as a guideline for the mesh selection. > Lateral flows are mainly exiting through upper and lower windows of guide tubes ({approx}81%) and 18% flow through small side gaps. > The interactions between guide tubes and neighboring support column as well as flow characteristic are revealed. - Abstract: One aspect of the Westinghouse AP1000{sup TM} reactor design is the reduction in the number of major components and simplification in manufacturing. One design change relative to current Westinghouse reactors of similar size is that AP1000 reactor vessel has two nozzles/hot legs instead of three. With regard to fuel performance, this design difference creates a different flow field in the reactor vessel upper plenum. The flow exiting from the core and entering the upper plenum must turn toward one of the two outlet nozzles and flow laterally around numerous control rod guide tubes and support columns. Also, below the upper plenum are the upper core plate and the top core region of the 157 fuel assemblies and 69 guidetube assemblies. To determine how the lateral flow in the top of the core and upper plenum compares to the current reactors a CFD model of the flow in the upper portion of the AP1000 reactor vessel was created. Before detailed CFD simulations of the flow in the entire upper plenum and top core regions were performed, conducting local simulations for smaller sections of the domain provided crucial and detailed physical aspects of the flow. These sub-domain models were used to perform mesh sensitivities and to assess what geometrical details may be eliminated from the larger model in order to reduce mesh size and computational requirements. In this paper

  13. ADDITIONAL STRESS AND FRACTURE MECHANICS ANALYSES OF PRESSURIZED WATER REACTOR PRESSURE VESSEL NOZZLES

    Energy Technology Data Exchange (ETDEWEB)

    Walter, Matthew [Structural Integrity Associates, Inc.; Yin, Shengjun [ORNL; Stevens, Gary [U.S. Nuclear Regulatory Commission; Sommerville, Daniel [Structural Integrity Associates, Inc.; Palm, Nathan [Westinghouse Electric Company, Cranberry Township, PA; Heinecke, Carol [Westinghouse Electric Company, Cranberry Township, PA

    2012-01-01

    In past years, the authors have undertaken various studies of nozzles in both boiling water reactors (BWRs) and pressurized water reactors (PWRs) located in the reactor pressure vessel (RPV) adjacent to the core beltline region. Those studies described stress and fracture mechanics analyses performed to assess various RPV nozzle geometries, which were selected based on their proximity to the core beltline region, i.e., those nozzle configurations that are located close enough to the core region such that they may receive sufficient fluence prior to end-of-life (EOL) to require evaluation of embrittlement as part of the RPV analyses associated with pressure-temperature (P-T) limits. In this paper, additional stress and fracture analyses are summarized that were performed for additional PWR nozzles with the following objectives: To expand the population of PWR nozzle configurations evaluated, which was limited in the previous work to just two nozzles (one inlet and one outlet nozzle). To model and understand differences in stress results obtained for an internal pressure load case using a two-dimensional (2-D) axi-symmetric finite element model (FEM) vs. a three-dimensional (3-D) FEM for these PWR nozzles. In particular, the ovalization (stress concentration) effect of two intersecting cylinders, which is typical of RPV nozzle configurations, was investigated. To investigate the applicability of previously recommended linear elastic fracture mechanics (LEFM) hand solutions for calculating the Mode I stress intensity factor for a postulated nozzle corner crack for pressure loading for these PWR nozzles. These analyses were performed to further expand earlier work completed to support potential revision and refinement of Title 10 to the U.S. Code of Federal Regulations (CFR), Part 50, Appendix G, Fracture Toughness Requirements, and are intended to supplement similar evaluation of nozzles presented at the 2008, 2009, and 2011 Pressure Vessels and Piping (PVP

  14. ASSESSMENT OF THE POTENTIAL FOR HYDROGEN GENERATION DURING GROUTING OPERATIONS IN THE R- AND P-REACTOR VESSELS

    Energy Technology Data Exchange (ETDEWEB)

    Wiersma, B.

    2009-12-29

    The R- and P-reactor buildings were retired from service and are now being prepared for deactivation and decommissioning (D&D). D&D activities will consist primarily of immobilizing contaminated components and structures in a grout-like formulation. Aluminum corrodes very rapidly when it comes in contact with the alkaline grout materials and as a result produces hydrogen gas. To address this potential deflagration/explosion hazard, the Materials Science and Technology Directorate (MS&T) of the Savannah River National Laboratory (SRNL) has been requested to review and evaluate existing experimental and analytical studies of this issue to determine if any process constraints on the chemistry of the fill material and the fill operation are necessary. Various options exist for the type of grout material that may be used for D&D of the reactor vessels. The grout formulation options include ceramicrete (pH 6-8), low pH portland cement + silica fume grout (pH 10.4), or Portland cement grout (pH 12.5). The assessment concluded that either ceramicrete or the silica fume grout may be used to safely grout the P-reactor vessel. The risk of accumulation of a flammable mixture of hydrogen between the grout-air interface and the top of the reactor is very low. Portland cement grout, on the other hand, for the same range of process parameters does not provide a significant margin of safety against the accumulation of flammable gas in the reactor vessel during grouting operations in the P-reactor vessel. It is recommended that this grout not be utilized for this task. The R-reactor vessel contains significantly less aluminum and thus a Portland cement grout may be considered as well. For example, if the grout fill rate is less than 1 inch/min and the grout temperature is maintained at 70 C or less, the risk of hydrogen accumulation in the R-reactor vessel is very low for the Portland cement. Alternatively, if the grout fill rate is less than 0.5 inch/min and the grout is maintained

  15. 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)

  16. On the effect of temperature on the threshold stress intensity factor of delayed hydride cracking in light water reactor fuel cladding

    Energy Technology Data Exchange (ETDEWEB)

    Holston, Anna-MariaAlvarez; Stjarnsater, Johan [Studsvik Nuclear AB, Nykoping (Sweden)

    2017-06-15

    Delayed hydride cracking (DHC) was first observed in pressure tubes in Canadian CANDU reactors. In light water reactors, DHC was not observed until the late 1990s in high-burnup boiling water reactor (BWR) fuel cladding. In recent years, the focus on DHC has resurfaced in light of the increased interest in the cladding integrity during interim conditions. In principle, all spent fuel in the wet pools has sufficient hydrogen content for DHC to operate below 300°C. It is therefore of importance to establish the critical parameters for DHC to operate. This work studies the threshold stress intensity factor (K{sub IH}) to initiate DHC as a function of temperature in Zry-4 for temperatures between 227°C and 315°C. The experimental technique used in this study was the pin-loading testing technique. To determine the K{sub IH}, an unloading method was used where the load was successively reduced in a stepwise manner until no cracking was observed during 24 hours. The results showed that there was moderate temperature behavior at lower temperatures. Around 300°C, there was a sharp increase in K{sub IH} indicating the upper temperature limit for DHC. The value for K{sub IH} at 227°C was determined to be 2.6 ± 0.3 MPa √m.

  17. Deformation behavior in reactor pressure vessel steels as a clue to understanding irradiation hardening.

    Energy Technology Data Exchange (ETDEWEB)

    DiMelfi, R. J.; Alexander, D. E.; Rehn, L. E.

    1999-10-25

    In this paper, we examine the post-yield true stress vs true strain behavior of irradiated pressure vessel steels and iron-based alloys to reveal differences in strain-hardening behavior associated with different irradiating particles (neutrons and electrons) and different alloy chernky. It is important to understand the effects on mechanical properties caused by displacement producing radiation of nuclear reactor pressure steels. Critical embrittling effects, e.g. increases in the ductile-to-brittle-transition-temperature, are associated with irradiation-induced increases in yield strength. In addition, fatigue-life and loading-rate effects on fracture can be related to the post-irradiation strain-hardening behavior of the steels. All of these properties affect the expected service life of nuclear reactor pressure vessels. We address the characteristics of two general strengthening effects that we believe are relevant to the differing defect cluster characters produced by neutrons and electrons in four different alloys: two pressure vessel steels, A212B and A350, and two binary alloys, Fe-0.28 wt%Cu and Fe-0.74 wt%Ni. Our results show that there are differences in the post-irradiation mechanical behavior for the two kinds of irradiation and that the differences are related both to differences in damage produced and alloy chemistry. We find that while electron and neutron irradiations (at T {le} 60 C) of pressure vessel steels and binary iron-based model alloys produce similar increases in yield strength for the same dose level, they do not result in the same post-yield hardening behavior. For neutron irradiation, the true stress flow curves of the irradiated material can be made to superimpose on that of the unirradiated material, when the former are shifted appropriately along the strain axis. This behavior suggests that neutron irradiation hardening has the same effect as strain hardening for all of the materials analyzed. For electron irradiated steels, the

  18. 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

  19. 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)

  20. Statistical analysis using the Bayesian nonparametric method for irradiation embrittlement of reactor pressure vessels

    Science.gov (United States)

    Takamizawa, Hisashi; Itoh, Hiroto; Nishiyama, Yutaka

    2016-10-01

    In order to understand neutron irradiation embrittlement in high fluence regions, statistical analysis using the Bayesian nonparametric (BNP) method was performed for the Japanese surveillance and material test reactor irradiation database. The BNP method is essentially expressed as an infinite summation of normal distributions, with input data being subdivided into clusters with identical statistical parameters, such as mean and standard deviation, for each cluster to estimate shifts in ductile-to-brittle transition temperature (DBTT). The clusters typically depend on chemical compositions, irradiation conditions, and the irradiation embrittlement. Specific variables contributing to the irradiation embrittlement include the content of Cu, Ni, P, Si, and Mn in the pressure vessel steels, neutron flux, neutron fluence, and irradiation temperatures. It was found that the measured shifts of DBTT correlated well with the calculated ones. Data associated with the same materials were subdivided into the same clusters even if neutron fluences were increased.

  1. On flux effects in a low alloy steel from a Swedish reactor pressure vessel

    Science.gov (United States)

    Boåsen, Magnus; Efsing, Pål; Ehrnstén, Ulla

    2017-02-01

    This study aims to investigate the presence of Unstable Matrix Defects in irradiated pressure vessel steel from weldments of the Swedish PWR Ringhals 4 (R4). Hardness tests have been performed on low flux (surveillance material) and high flux (Halden reactor) irradiated material samples in combination with heat treatments at temperatures of 330, 360 and 390 °C in order to reveal eventual recovery of any hardening features induced by irradiation. The experiments carried out in this study could not reveal any hardness recovery related to Unstable Matrix Defects at relevant temperatures. However, a difference in hardness recovery was found between the low and the high flux samples at heat treatments at higher temperatures than expected for the annihilation of Unstable Matrix Defects-the observed recovery is here attributed to differences of the solute clusters formed by the high and low flux irradiations.

  2. Atomic and dislocation dynamics simulations of plastic deformation in reactor pressure vessel steel

    Science.gov (United States)

    Monnet, Ghiath; Domain, Christophe; Queyreau, Sylvain; Naamane, Sanae; Devincre, Benoit

    2009-11-01

    The collective behavior of dislocations in reactor pressure vessel (RPV) steel involves dislocation properties on different phenomenological scales. In the multiscale approach, adopted in this work, we use atomic simulations to provide input data for larger scale simulations. We show in this paper how first-principles calculations can be used to describe the Peierls potential of screw dislocations, allowing for the validation of the empirical interatomic potential used in molecular dynamics simulations. The latter are used to compute the velocity of dislocations as a function of the applied stress and the temperature. The mobility laws obtained in this way are employed in dislocation dynamics simulations in order to predict properties of plastic flow, namely dislocation-dislocation interactions and dislocation interactions with carbides at low and high temperature.

  3. United States Department of Energy projects related to reactor pressure vessel annealing optimization

    Energy Technology Data Exchange (ETDEWEB)

    Rosinski, S.T.; Nakos, J.T.

    1993-09-01

    Light water reactor pressure vessel (RPV) material properties reduced by long-term exposure to neutron irradiation can be recovered through a thermal annealing treatment. This technique to extend RPV life, discussed in this report, provides a complementary approach to analytical methodologies to evaluate RPV integrity. RPV annealing has been successfully demonstrated in the former Soviet Union and on a limited basis by the US (military applications only). The process of demonstrating the technical feasibility of annealing commercial US RPVs is being pursued through a cooperative effort between the nuclear industry and the US Department of Energy (USDOE) Plant Lifetime Improvement (PLIM) Program. Presently, two projects are under way through the USDOE PLIM Program to demonstrate the technical feasibility of annealing commercial US RPVS, (1) annealing re-embrittlement data base development and (2) heat transfer boundary condition experiments.

  4. Development of an ultrasonic imaging system for the inspection of nuclear reactor pressure vessels. Final report

    Energy Technology Data Exchange (ETDEWEB)

    Becker, F.L.; Crow, V.L.; Davis, T.J.; Doctor, S.R.; Hildebrand, B.P.; Lemon, D.K.; Posakony, G.J.

    1979-10-01

    The development of an experimental model of an ultrasonic linear array system for the inspection of weldments in nuclear reactor pressure vessels is described. The imaging system is designed to operate in both pulse echo and holographic modes of operation. The system utilizes a sequentially pulsed, phase steered linear array to develop pulse echo images and a line focused illumination transducer in conjunction with a linear receiver array to develop holographic reconstructed images. The results recorded from the computer-based system demonstrate the capability of array technology. Excellent results from both the pulse echo and holographic modes of operation have been achieved. Pulse echo images of flaws in weldments are displayed in B-scan, C-scan, or isometric presentations. Reconstruction of the phase or holographic images are compared with pulse echo results and demonstrate the enhancement potential for the holographic procedure.

  5. Qualification of non-destructive examination for belgian nuclear reactor pressure vessel inspection

    Energy Technology Data Exchange (ETDEWEB)

    Couplet, D. [TRACTEBEL, Brussels (Belgium); Francoise, T. [Intercontrole, 94 - Rungis (France)

    2001-07-01

    In Service Inspection (ISI) participates to the assessment of Nuclear Reactor Pressure Vessel Integrity. The performance of Non Destructive Examination (NDE) techniques must be demonstrated according to predefined objectives. The qualification process is essential to trust the reliability of the information provided by NDE. In Belgian Nuclear Power Plants, the qualification was conducted through a collaboration between the vendor and a technical group from the Electricity Utility. The important facts of this qualification will be presented: - the detailed definition of the inspection and qualifications objectives, based on a combination of the ASME code and the European Methodology for Qualification; - the systematic verification of the NDE performance and limitations, for each ISI objective, through an adequate combination of tests on blocks and technical justification; - the continuous improvement of the NDE procedure; - the feedback and the lessons learnt from site experience; - the necessary multi-disciplinary approach (NDE, degradation mechanisms, structural integrity)

  6. Microstructural investigations on Russian reactor pressure vessel steels by small-angle neutron scattering

    Science.gov (United States)

    Ulbricht, A.; Boehmert, J.; Strunz, P.; Dewhurst, C.; Mathon, M.-H.

    The effect of radiation embrittlement has a high safety significance for Russian VVER reactor pressure vessel steels. Heats of base and weld metals of the as-received state, irradiated state and post-irradiation annealed state were investigated using small-angle neutron scattering (SANS) to obtain insight about the microstructural features caused by fast neutron irradiation. The SANS intensities increase in the momentum transfer range between 0.8 and 3 nm-1 for all the material compositions in the irradiated state. The size distribution function of the irradiation-induced defect clusters has a pronounced maximum at 1 nm in radius. Their content varies between 0.1 and 0.7 vol.% dependent on material composition and increases with the neutron fluence. The comparison of nuclear and magnetic scattering indicates that the defects differ in their composition. Thermal annealing reduces the volume fraction of irradiation defect clusters.

  7. Reactor pressure vessel integrity research at the Oak Ridge National Laboratory

    Energy Technology Data Exchange (ETDEWEB)

    Corwin, W.R.; Pennell, W.E.; Pace, J.V.

    1995-12-31

    Maintaining the integrity of the reactor pressure vessel (RPV) in a light-water-cooled nuclear power plant is crucial in preventing and controlling severe accidents that have the potential for major contamination release. The RPV is the only key safety-related component of the plant for which a duplicate or redundant backup system does not exist. It is therefore imperative to understand and be able to predict the integrity inherent in the RPV. For this reason, the U.S. Nuclear Regulatory Commission has established the related research programs at ORNL described herein to provide for the development and confirmation of the methods used for: (1) establishing the irradiation exposure conditions within the RPV in the Embrittlement Data Base and Dosimetry Evaluation Program, (2) assessing the effects of irradiation on the RPV materials in the Heavy-Section Steel Irradiation Program, and (3) developing overall structural and fracture analyses of RPVs in the Heavy-Section Steel Technology Program.

  8. Atomic and dislocation dynamics simulations of plastic deformation in reactor pressure vessel steel

    Energy Technology Data Exchange (ETDEWEB)

    Monnet, Ghiath, E-mail: ghiathmonnet@yahoo.f [EDF-R and D, MMC, Avenue des Renardieres, 77818 Moret sur Loing (France); Domain, Christophe; Queyreau, Sylvain; Naamane, Sanae [EDF-R and D, MMC, Avenue des Renardieres, 77818 Moret sur Loing (France); Devincre, Benoit [LEM, CNRS-ONERA, 29 av. de la division Leclerc, 92130 Chatillon (France)

    2009-11-15

    The collective behavior of dislocations in reactor pressure vessel (RPV) steel involves dislocation properties on different phenomenological scales. In the multiscale approach, adopted in this work, we use atomic simulations to provide input data for larger scale simulations. We show in this paper how first-principles calculations can be used to describe the Peierls potential of screw dislocations, allowing for the validation of the empirical interatomic potential used in molecular dynamics simulations. The latter are used to compute the velocity of dislocations as a function of the applied stress and the temperature. The mobility laws obtained in this way are employed in dislocation dynamics simulations in order to predict properties of plastic flow, namely dislocation-dislocation interactions and dislocation interactions with carbides at low and high temperature.

  9. The results and analysis of irradiation experiments conducted on reactor vessel plate and weld materials

    Energy Technology Data Exchange (ETDEWEB)

    Biemiller, E.C. [Yankee Atomic Electric Co., Bolton, MA (United States); Carter, R.G.; Rosinski, S.T. [Electric Power Research Inst., Charlotte, NC (United States)

    1996-09-01

    This paper documents the extensive amount of experimental work on radiation damage to reactor vessel materials carried out by Yankee Atomic Electric Company (YAEC) and others in support of a licensing effort to restart the Yankee Rowe nuclear power plant. The effect of plate nickel content and microstructure on irradiation damage sensitivity was assessed. Typical reactor pressure vessel plate materials each containing 0.24% (by weight) copper, but different nickel contents at 0.19% and 0.63% were heat treated to produce different microstructures. A Linde 80 weld containing 0.30% copper and 1.00% nickel was produced and heat treated to test microstructure effects on the irradiation response of weld metal. Materials taken from plate surface locations (vs 1/4%) were included to test whether or not the improved toughness properties of the plate surface layer, resulting from a rapid quench, is maintained after irradiation. Irradiations were conducted at two irradiation temperatures, 500 F (260 C) and 550 F (288 C), to determine the effect of irradiation temperature on embrittlement. The results of this irradiation testing and additional data from a DOE/Sandia National Laboratories irradiation study show an irradiation temperature effect that is not consistent, but varies with the materials tested. The test results demonstrate that for nickel bearing steels, the superior toughness of plate surface material is maintained even after irradiation to high fluences, and for the copper content tested, nickel has little effect on irradiation response. A mixed effect of microstructure/heat treatment on the materials` irradiation response was noted. Phosphorus potentially played a role in the irradiation response of the low nickel material irradiated at 500 F (288 C) but did not show prominence in the irradiations for the same material conducted at 500 F (260 C).

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

    Energy Technology Data Exchange (ETDEWEB)

    Yun, Sunghwan; Kim, Sang Ji [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2015-10-15

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

  11. Effects of thermal annealing and reirradiation on toughness of reactor pressure vessel steels

    Energy Technology Data Exchange (ETDEWEB)

    Nanstad, R.K.; Iskander, S.K.; Sokolov, M.A. [Oak Ridge National Lab., TN (United States)] [and others

    1997-02-01

    One of the options to mitigate the effects of irradiation on reactor pressure vessels (RPV) is to thermally anneal them to restore the toughness properties that have been degraded by neutron irradiation. This paper summarizes recent experimental results from work performed at the Oak Ridge National Laboratory (ORNL) to study the annealing response, or {open_quotes}recovery,{close_quotes} of several irradiated RPV steels; it also includes recent results from both ORNL and the Russian Research Center-Kurchatov Institute (RRC-KI) on a cooperative program of irradiation, annealing and reirradiation of both U.S. and Russian RPV steels. The cooperative program was conducted under the auspices of Working Group 3, U.S./Russia Joint Coordinating Committee for Civilian Nuclear Reactor Safety (JCCCNRS). The materials investigated are an RPV plate and various submerged-arc welds, with tensile, Charpy impact toughness, and fracture toughness results variously determined. Experimental results are compared with applicable prediction guidelines, while observed differences in annealing responses and reirradiation rates are discussed.

  12. Residual Stress Evaluation of Weld Inlay Process on Reactor Vessel Nozzles

    Energy Technology Data Exchange (ETDEWEB)

    Cho, Kihyun; Cho, Hong Seok [KEPCO KPS, Naju (Korea, Republic of)

    2015-10-15

    Weld overlay, weld inlay and stress improvement are mitigation technologies for butt joints. Weld overlay is done on pressurizer nozzles which are the highest potential locations occurring PWSCC due to high temperature in Korea. Reactor vessel nozzles are other big safety concerns for butt joints. Weld overlay and stress improvement should be so difficult to apply to those locations because space is too limited. Weld inlay should be one of the solutions. KEPCO KPS has developed laser welding system and process for reactor nozzles. Welding residual stress analysis is necessary for flaw evaluation. United States nuclear regulatory commission has calculated GTAW(Gas Tungsten Arc Welding) residual stress using ABAQUS. To confirm effectiveness of weld inlay process, welding residual stress analysis was performed. and difference between GTAW and LASER welding process was compared. Evaluation of weld inlay process using ANSYS and ABAQUS is performed. All of the both results are similar. The residual stress generated after weld inlay was on range of 450-500 MPa. Welding residual stresses are differently generated by GTAW and LASER welding. But regardless of welding process type, residual tensile stress is generated on inside surface.

  13. Standard Master Matrix for Light-Water Reactor Pressure Vessel Surveillance Standards, E706(0)

    CERN Document Server

    American Society for Testing and Materials. Philadelphia

    2002-01-01

    1.1 This 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 (PV) and support structure steels throughout a pressure vessel's service life (Fig. 1). Some of these are existing ASTM standards, some are ASTM standards that have been modified, and some are proposed ASTM standards. General requirements of content and consistency are discussed in Section 6 . More detailed writers' and users' information, justification, and specific requirements for the nine practices, ten guides, and three methods are provided in Sections 3-5. Referenced documents are discussed in Section 2. The summary-type information that is provided in Sections 3 and 4 is essential for establishing proper understanding and communications between the writers and users of this set of matrix standards. It was extracted from the referenced documents, Section 2 and references (1-106) for use by individual writers and users. 1...

  14. Current Status of Development of High Nickel Low Alloy Steels for Commercial Reactor Pressure Vessel

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Min Chul; Lee, B. S.; Park, S. G.; Lee, K. H. [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2008-12-15

    SA508 Gr.3 Mn-Mo-Ni low alloy steels have been used for nuclear reactor pressure vessel steels up to now. Currently, the design goal of nuclear power plant is focusing at larger capacity and longer lifetime. Requirements of much bigger pressure vessels may cause critical problems in the manufacturing stage as well as for the welding stage. Application of higher strength steel may be required to overcome the technical problems. It is known that a higher strength and fracture toughness of low alloy steels such as SA508 Gr.4N low alloy steel could be achieved by increasing the Ni and Cr contents. Therefore, SA508 Gr.4N low alloy steel is very attractive as eligible RPV steel for the next generation PWR systems. In this report, we propose the possibility of SA508 Gr.4N low alloy steel for an application of next generation commercial RPV, based on the literature research result about development history of the RPV steels and SA508 specification. In addition, we have surveyed the research result of HSLA(High Strength Low Alloy steel), which has similar chemical compositions with SA508 Gr.4N, to understand the problems and the way of improvement of SA508 Gr.4N low alloy steel. And also, we have investigated eastern RPV steel(WWER-1000), which has higher Ni contents compared to western RPV steel.

  15. Development of in-vessel type control rod drive mechanism for a innovative small reactor (Contract research)

    Energy Technology Data Exchange (ETDEWEB)

    Yoritsune, Tsutomu; Ishida, Toshihisa [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    2003-03-01

    Although the control rod drive mechanism of an existing large scale light water reactor is generally installed outside the reactor vessel, an in-vessel type control rod drive mechanism (INV-CRDM) is installed inside the reactor vessel. The INV-CRDM contributes to compactness and simplicity of the reactor system, and it can eliminate the possibility of a rod ejection accident. Therefore, INV-CRDM is an important technology adopted in an innovative small reactor. Japan Atomic Energy Research Institute (JAERI) has developed this type of CRDM driven by an electric motor, which can work under high temperature and high pressure water for the advanced marine reactor. On the basis of this research result, a driving motor coil and a bearing were developed to be used under the high temperature steam, severe condition for an innovative small reactor. About the driving motor, we manufactured the driving motor available for high temperature steam and carried out performance test under room temperature atmosphere to confirm the electric characteristic and coolability of the driving coil. With these test results and the past test results under high temperature water, we analyzed and evaluated the electric performance and coolability of the driving coil under high temperature steam. Concerning bearing, we manufactured the test pieces using some candidate material for material characteristic test and carried out the rolling wear test under high temperature steam to select the material. Consequently, we confirmed that performance of the driving coil for the advanced type driving motor, is enough to be used under high temperature steam. And, we evaluated the performance of the bearing and selected the material of the bearing, which can be used under high temperature steam. From these results, we have obtained the prospect that the INV-CRDM can be used for an innovative small reactor under steam atmosphere could be developed. (author)

  16. Development of in-vessel type control rod drive mechanism for a innovative small reactor (Contract research)

    Energy Technology Data Exchange (ETDEWEB)

    Yoritsune, Tsutomu; Ishida, Toshihisa [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    2003-03-01

    Although the control rod drive mechanism of an existing large scale light water reactor is generally installed outside the reactor vessel, an in-vessel type control rod drive mechanism (INV-CRDM) is installed inside the reactor vessel. The INV-CRDM contributes to compactness and simplicity of the reactor system, and it can eliminate the possibility of a rod ejection accident. Therefore, INV-CRDM is an important technology adopted in an innovative small reactor. Japan Atomic Energy Research Institute (JAERI) has developed this type of CRDM driven by an electric motor, which can work under high temperature and high pressure water for the advanced marine reactor. On the basis of this research result, a driving motor coil and a bearing were developed to be used under the high temperature steam, severe condition for an innovative small reactor. About the driving motor, we manufactured the driving motor available for high temperature steam and carried out performance test under room temperature atmosphere to confirm the electric characteristic and coolability of the driving coil. With these test results and the past test results under high temperature water, we analyzed and evaluated the electric performance and coolability of the driving coil under high temperature steam. Concerning bearing, we manufactured the test pieces using some candidate material for material characteristic test and carried out the rolling wear test under high temperature steam to select the material. Consequently, we confirmed that performance of the driving coil for the advanced type driving motor, is enough to be used under high temperature steam. And, we evaluated the performance of the bearing and selected the material of the bearing, which can be used under high temperature steam. From these results, we have obtained the prospect that the INV-CRDM can be used for an innovative small reactor under steam atmosphere could be developed. (author)

  17. Development of Reactor Vessel Bottom Mount Instrumentation Nozzle Routine Inspection Device

    Energy Technology Data Exchange (ETDEWEB)

    Khaled, Atya Ahmed Abdallah; Ihn, Namgung [KEPCO International Nuclear Graduate School, Ulsan (Korea, Republic of)

    2015-10-15

    The primary coolant water of pressurized water reactors has created cracks in j-weld of penetrations with Alloy 600 through a process called primary water stress corrosion cracking. On October 6, 2013, BMI nozzle number 3 at Palo Verde Unit 3 (PVNGS-3) exhibited small white de-posits around the annulus. Nozzle attachment to the RV lower head is by J-groove weld to the inside penetration of the nozzle and the weld material is of Alloy 600 material. Above two cases clearly show the necessity of routine inspection of RV lower head penetration during refueling outage. Nondestructive inspection is generally performed to detect fine cracks or defects that may develop during operation. Defects usually occur at weld regions, hence most non-destructive inspection is to scan and check any defects or crack in the weld region. BMI nozzles at the bottom head of a nuclear reactor vessel (RV) are one of such area for inspection. But BMI nozzles have not been inspected during regular refuel outage due to the relative small size of BMI nozzle and limited impact of the consequences of BMI leak. However, there is growing concern since there have been leaks at nuclear power plants (NPPs) as well as recent operating experience. In this study, we propose a system that is conveniently used for nondestructive inspection of BMI nozzles during regular refueling outage without removing all the reactor internals. A 3D model of the inspection system was also developed along with the RV and internals which permits a virtual 3D simulation to check the design concept and usability of the system.

  18. Initial Probabilistic Evaluation of Reactor Pressure Vessel Fracture with Grizzly and Raven

    Energy Technology Data Exchange (ETDEWEB)

    Spencer, Benjamin [Idaho National Lab. (INL), Idaho Falls, ID (United States); Hoffman, William [Univ. of Idaho, Moscow, ID (United States); Sen, Sonat [Idaho National Lab. (INL), Idaho Falls, ID (United States); Rabiti, Cristian [Idaho National Lab. (INL), Idaho Falls, ID (United States); Dickson, Terry [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Bass, Richard [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2015-10-01

    The Grizzly code is being developed with the goal of creating a general tool that can be applied to study a variety of degradation mechanisms in nuclear power plant components. The first application of Grizzly has been to study fracture in embrittled reactor pressure vessels (RPVs). Grizzly can be used to model the thermal/mechanical response of an RPV under transient conditions that would be observed in a pressurized thermal shock (PTS) scenario. The global response of the vessel provides boundary conditions for local models of the material in the vicinity of a flaw. Fracture domain integrals are computed to obtain stress intensity factors, which can in turn be used to assess whether a fracture would initiate at a pre-existing flaw. These capabilities have been demonstrated previously. A typical RPV is likely to contain a large population of pre-existing flaws introduced during the manufacturing process. This flaw population is characterized stastistically through probability density functions of the flaw distributions. The use of probabilistic techniques is necessary to assess the likelihood of crack initiation during a transient event. This report documents initial work to perform probabilistic analysis of RPV fracture during a PTS event using a combination of the RAVEN risk analysis code and Grizzly. This work is limited in scope, considering only a single flaw with deterministic geometry, but with uncertainty introduced in the parameters that influence fracture toughness. These results are benchmarked against equivalent models run in the FAVOR code. When fully developed, the RAVEN/Grizzly methodology for modeling probabilistic fracture in RPVs will provide a general capability that can be used to consider a wider variety of vessel and flaw conditions that are difficult to consider with current tools. In addition, this will provide access to advanced probabilistic techniques provided by RAVEN, including adaptive sampling and parallelism, which can dramatically

  19. Heating of reactor pressure vessel bottom head and penetrations in a severe reactor accident; Reaktoripaineastian pohjan ja laepivientien kuumeneminen sydaemen sulamisonnettomuudessa

    Energy Technology Data Exchange (ETDEWEB)

    Ikonen, K. [VTT Energy, Espoo (Finland). Nuclear Energy

    1997-10-01

    The report describes the fundamentals of heat conductivity and convection and numerical methods like finite difference and control volume method for calculation of the thermal history of a reactor pressure vessel bottom head and penetrations. Phase changes from solids to liquids are considered. Time integration is performed by explicit or implicit method. Developed computer codes for thermal conductivity and convection analyses and codes for graphical visualization are described. The codes are applied to two practical cases. They deal with analyses of Swiss CORVIS-experiments and analyses of control rod and instrument penetrations in a BWR bottom head. A model for calculation of effective thermal conductivity of granular corium is developed. The work is also related to EU MVI-project (Core Melt-Pressure Vessel Interactions During a Light Water Reactor Severe Accident), whose coordinator is Prof. B. R. Sehgal at Royal Institute of Technology in Stockholm. (orig.) (11 refs.).

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

    Energy Technology Data Exchange (ETDEWEB)

    Langton, C.; Stefanko, D.

    2011-03-10

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

  1. Reactor Pressure Vessel Integrity Assessments with the Grizzly Aging Simulation Code

    Energy Technology Data Exchange (ETDEWEB)

    Spencer, Benjamin; Backman, Marie; Hoffman, William; Chakraborty, Pritam

    2015-08-01

    Grizzly is a simulation tool being developed at Idaho National Laboratory (INL) as part of the US Department of Energy’s Light Water Reactor Sustainability program to provide improved safety assessments of systems, components, and structures in nuclear power plants subjected to age-related degradation. Its goal is to provide an improved scientific basis for decisions surrounding license renewal, which would permit operation of commercial nuclear power plants beyond 60 years. Grizzly is based on INL’s MOOSE framework, which enables multiphysics simulations in a parallel computing environment. It will address a wide variety of aging issues in nuclear power plant systems, components, and structures, modelling both the aging processes and the ability of age-degraded components to perform safely. The reactor pressure vessel (RPV) was chosen as the initial application for Grizzly. Grizzly solves tightly coupled equations of heat conduction and solid mechanics to simulate the global response of the RPV to accident conditions, and uses submodels to represent regions with pre-existing flaws. Domain integrals are used to calculate stress intensity factors on those flaws. A physically based empirical model is used to evaluate material embrittlement, and is used to evaluate whether crack growth would occur. Grizzly can represent the RPV in 2D or 3D, allowing it to evaluate effects that require higher dimensionality models to capture. Work is underway to use lower length scale models of material evolution to inform engineering models of embrittlement. This paper demonstrates an application of Grizzly to RPV failure assessment, and summarizes on-going work.

  2. Comparison of microstructural features of radiation embrittlement of VVER-440 and VVER-1000 reactor pressure vessel steels

    Science.gov (United States)

    Kuleshova, E. A.; Gurovich, B. A.; Shtrombakh, Ya. I.; Erak, D. Yu.; Lavrenchuk, O. V.

    2002-02-01

    Comparative microstructural studies of both surveillance specimens and reactor pressure vessel (RPV) materials of VVER-440 and VVER-1000 light water reactor systems have been carried out, following irradiation to different fast neutron fluences and of the heat treatment for extended periods at the operating temperatures. It is shown that there are several microstructural features in the radiation embrittlement of VVER-1000 steels compared to VVER-440 RPV steels that can cause changes in the contributions of different radiation embrittlement mechanisms for VVER-1000 steel.

  3. Crack growth tests on a ferritic reactor pressure vessel steel under the simultaneous influence of simulated BWR coolant and irradiation

    Energy Technology Data Exchange (ETDEWEB)

    Hoffmann, H. [VGB PowerTech e.V., Essen (Germany); Huettner, F. [Hamburgische Electricitaets-Werke AG, Hamburg (Germany); Ilg, U. [EnBW Kraftwerke AG, Philippsburg (Germany); Wachter, O. [E.ON Kernkraft GmbH, Hannover(Germany); Widera, M. [RWE Power AG, Essen (Germany); Brozova, A.; Ernestova, M.; Kysela, J.; Vsolak, R. [Nuclear Research Institute Rez plc (Czech Republic)

    2004-07-01

    Crack growth tests under constant load with initial in-situ cycling were performed on the low alloy reactor pressure vessel (RPV) steel 22 NiMoCr 3 7 (A 508 Cl. 2) with the goal to determine crack growth rates of irradiated and non-irradiated steel under the simultaneous influence of simulated BWR coolant and irradiation. The tests were performed under conditions as near as possible to operational conditions in a commercial BWR reactor. The research results are summarized and are compared with international data. (orig.)

  4. Integrity analysis of reactor pressure vessels subjected to pressurized thermal shocks by XFEM

    Energy Technology Data Exchange (ETDEWEB)

    González-Albuixech, V.F., E-mail: vicente.gonzalez@psi.ch; Qian, G.; Niffenegger, M.

    2014-08-15

    Highlights: • We did fracture mechanics computations for an RPV with XFEM thermal shocks. • We introduce guidelines for using XFEM in RPV studies. • We did a comparison between FEM and XFEM results for an RPV analysis. • Some limitations of the eXtended Finite Element Methods are commented. - Abstract: The integrity of an reactor pressure vessel (RPV) related to Pressurized Thermal Shocks (PTSs) has been widely studied. However, due to the difficulties associated with the crack modeling with the 3-D finite element method (FEM), it is preferred to use models with simple geometries and crack configurations. In the last years new improved FEMs were developed which include the singularities and discontinuities and simplify the computational fracture mechanics studies. One of those methods, the eXtended Finite Element Method (XFEM) relies on the introduction of the crack effect with an enrichment of the finite element approximation space. This paper introduces the use of XFEM to the structural analysis of an RPV subjected to PTSs. The analysis compares the stress intensity factor (SIF) calculated with XFEM with results obtained by conventional FEM calculations.

  5. Assessment of the advantages of a residual heat removal system inside the reactor pressure vessel

    Energy Technology Data Exchange (ETDEWEB)

    Gautier, G.M.

    1995-12-31

    In the framework of research on diversified means for removing the residual heat from pressurized water reactors, the CEA is studying a passive system called RRP (Refroidissement du Reacteur au Primaire, or primary circuit cooling system), which includes integrated heat-exchangers and a layout of the internal structures so as to obtain convection from the primary circuit inside the vessel, whatever the state of the loops. This system is operational for all primary circuit temperatures and pressures, as well as for a wide range of conditions: it is independent of the state of the loops, even if the volume of water in the primary circuit is small, it is compatible with either a passive or an active operation mode, and compatible with any other decay heat removal systems. An evaluation is presented here of the performance of the RRP system in the event of a small primary circuit break in a totally passive operation mode without the intervention of another system. The results of this evaluation show the interest of such a system: a clear increase of the time-delay for the implementation of a low pressure safety injection system, no need for the use of a high pressure safety injection system. (author). 4 refs., 7 figs., 1 tab.

  6. Analysis for the Effect of Spatial Discretization Method on AP1000 Reactor Pressure Vessel Fluence Calculation

    Directory of Open Access Journals (Sweden)

    Junxiao Zheng

    2016-01-01

    Full Text Available Maintaining the structural integrity of the reactor pressure vessel (RPV is a critical concern related to the safe operation of nuclear power plants. To estimate the structural integrity over the designed lifetime and to support analyses for a potential plant life extension, an accurate calculation of the fast neutron fluence (E>1.0 MeV or E>0.1 MeV at the RPV is significant. The discrete ordinates method is one of the main methods to solve such problems. During the calculation process, many factors will affect the results. In this paper, the deviations introduced by different differencing schemes and mesh sizes on the AP1000 RPV fast neutron fluence have been studied, which are based on new discrete ordinates code ARES. The analysis shows that the differencing scheme (diamond difference with or without linear zero fix-up, theta weighted, directional theta weighted, and exponential directional weighted introduces a deviation within 4%. The coarse mesh (4 × 4 cm meshes in XY plane leads to approximately 23.7% calculation deviation compared to those of refined mesh (1 × 1 cm meshes in XY plane. Comprehensive study on the deviation introduced by differencing scheme and mesh size has great significance for reasoned evaluation of RPV fast neutron fluence calculation results.

  7. Nondestructive characterization of embrittlement in reactor pressure vessel steels -- A feasibility study

    Energy Technology Data Exchange (ETDEWEB)

    McHenry, H.I.; Alers, G.A. [National Inst. of Standards and Technology, Boulder, CO (United States). Materials Reliability Div.

    1998-03-01

    The Nuclear Regulatory Commission recently initiated a study by NIST to assess the feasibility of using physical-property measurements for evaluating radiation embrittlement in reactor pressure vessel (RPV) steels. Ultrasonic and magnetic measurements provide the most promising approaches for nondestructive characterization of RPV steels because elastic waves and magnetic fields can sense the microstructural changes that embrittle materials. The microstructural changes of particular interest are copper precipitation hardening, which is the likely cause of radiation embrittlement in RPV steels, and the loss of dislocation mobility that is an attribute of the ductile-to-brittle transition. Measurements were made on a 1% copper steel, ASTM grade A710, in the annealed, peak-aged and overaged conditions, and on an RPV steel, ASTM grade A533B. Nonlinear ultrasonic and micromagnetic techniques were the most promising measures of precipitation hardening. Ultrasonic velocity measurements and the magnetic properties associated with hysteresis-loop measurements were not particularly sensitive to either precipitation hardening or the ductile-to-brittle transition. Measurements of internal friction using trapped ultrasonic resonance modes detected energy losses due to the motion of pinned dislocations; however, the ultrasonic attenuation associated with these measurements was small compared to the attenuation caused by beam spreading that would occur in conventional ultrasonic testing of RPVs.

  8. Fracture mechanics characterisation of the WWER-440 reactor pressure vessel beltline welding seam

    Energy Technology Data Exchange (ETDEWEB)

    Viehrig, Hans-Werner [Forschungszentrum Dresden-Rossendorf, P.O. Box 510119, 01314 Dresden (Germany)], E-mail: H.W.Viehrig@fzd.de; Schuhknecht, Jan [Forschungszentrum Dresden-Rossendorf, P.O. Box 510119, 01314 Dresden (Germany)

    2009-04-15

    The master curve (MC) approach as standardised in the ASTM Standard Test Method E1921 was applied to weld metal of the reactor pressure vessel (RPV) beltline welding seam of Greifswald unit 8 RPV. Charpy size SE(B) specimens from 13 locations equally spaced over the thickness of the welding seam were tested. The orientation of the specimens within the welding seam is TL and TS according to ASTM E399. The fracture toughness values measured on the SE(B) specimens with both orientations follow the course of the master curve. Nearly all values lie within the fracture toughness curves for 2% and 98% fracture probability. There is a strong variation of the reference temperature T{sub 0} through the thickness of the welding seam, which can be explained by microstructural differences. The scatter is more pronounced for the TS SE(B) specimens. It can be shown that specimens with TL and TS orientation in the welding seam have a differentiating and integrating behaviour, respectively.

  9. Effects of Microstructural Inhomogeneity on Charpy Impact Properties for Reactor Pressure Vessel

    Energy Technology Data Exchange (ETDEWEB)

    Hong, Seokmin; Song, Jaemin; Kim, Min-Chul; Choi, Kwon-Jae; Lee, Bong-Sang [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2014-10-15

    Reactor pressure vessel (RPV) steels are fabricated by vacuum carbon deoxidation (VCD), and then heat treatment of quenching and tempering is conducted after forging. The through-the-thickness variation of microstructure in RPV can occur due to the cooling rate gradient during quenching and inhomogeneous deformation during forging process. The variation of microstructure in RPV affects the mechanical properties, and inhomogeneity in mechanical properties can occur. The evaluation of mechanical properties of RPV is conducted at thickness of 1/4T. In order to evaluate the safety of RPV more correctly, the research about the through-the-thickness variation of microstructure and mechanical properties in RPV is need. 1. The fine low bainite (LB) is the dominant phase at the inner-surface (0T), but coarse upper bainite (UB) is the dominant phase at the center (1/2T). This is because cooling rate gradient from surface to center occurs during quenching. 2. Inter-lath carbides act as fracture initiation site, and it reduces impact toughness. 3. The upper shelf energy is low and the reference temperatures are high at the 1/4T. Impact properties are poor at 1/4T because of the formation of coarse upper bainite structure and coarse inter-lath carbides.

  10. Effect of tempering temperature on the microstructure and mechanical properties of a reactor pressure vessel steel

    Science.gov (United States)

    Li, C. W.; Han, L. Z.; Luo, X. M.; Liu, Q. D.; Gu, J. F.

    2016-08-01

    The microstructure and mechanical properties of reactor pressure vessel (RPV) steel were investigated after tempering at different temperatures ranging from 580 to 700 °C for 5 h. With increasing tempering temperature, the impact toughness, which is qualified by Charpy V-notch total absorbed energy, initially increases from 142 to 252 J, and then decreases to 47 J, with a maximum value at 650 °C, while the ultimate tensile strength varies in exactly the opposite direction. Comparing the microstructure and fracture surfaces of different specimens, the variations in toughness and strength with the tempering temperature were generally attributed to the softening of the bainitic ferrite, the agminated Fe3C carbides that resulted from decomposition of martensite/austenite (M/A) constituents, the precipitation of Mo2C carbides, and the newly formed M/A constituents at the grain boundaries. Finally, the correlation between the impact toughness and the volume fraction of the M/A constituents was established, and the fracture mechanisms for the different tempering conditions are explained.

  11. Views of TAGSI on the effects of gamma irradiation on the mechanical properties of irradiated ferritic steel reactor pressure vessels

    Energy Technology Data Exchange (ETDEWEB)

    Knott, J.F. [School of Engineering, Metallurgy and Materials, University of Birmingham, Edgbaston, Birmingham B15 2TT (United Kingdom); English, C.A. [Materials and Chemistry Consultancy, Nexia Solutions, 168 Harwell International Business Centre, Didcot, Oxon OX11 0QJ (United Kingdom); Weaver, D.R. [School of Physics, University of Birmingham, Edgbaston, Birmingham B15 2TT (United Kingdom); Lidbury, D.P.G. [Serco Assurance, Walton House, 404 The Quadrant, Birchwood Park, Warrington, Cheshire WA3 6AT (United Kingdom)

    2005-12-01

    The paper reviews and analyses the effects of gamma irradiation dose on the properties of ferritic steels used in reactor pressure vessels (RPVs). It explains factors that affect the embrittlement of a RPV steel induced by combinations of fast neutrons, thermal neutrons, and gamma irradiation. TAGSI were asked to consider the effects of gamma irradiation dose on the properties of steels used in reactor pressure vessels. TAGSI endorsed the use of the MCBEND code to calculate gamma fluxes and energetic gamma ray displacement cross-sections calculated using either Baumann or Alexander methods. TAGSI endorsed the calculation of the materials property changes due to an additional gamma dose using trend curves based on empirical correlation to neutron-induced damage (where k {sub {gamma}}{approx}1{+-}0.25)

  12. In vessel detection of delayed neutron emitters from clad failure in sodium cooled nuclear reactors: An estimation of the signal

    Science.gov (United States)

    Filliatre, P.; Jammes, C.; Chapoutier, N.; Jeannot, J.-P.; Jadot, F.; Batail, R.; Verrier, D.

    2014-04-01

    The detection of clad failures is mandatory in sodium-cooled fast neutron reactors in compliance with the "clean sodium" concept. An in-vessel detection system, sensitive to delayed neutrons from fission products released into the primary coolant by failures, partially tested in SUPERPHENIX, is foreseen in current SFR projects in order to reduce significantly the delay before an alarm is issued. In this paper, an estimation of the signal received by such a system in case of a failure is derived, taking the French project ASTRID as a working example. This failure induced signal is compared to that of the contribution of the neutrons from the core itself. The sensitivity of the system is defined in terms of minimal detectable surface of clad failure. Possible solutions to improve this sensitivity are discussed, involving either the sensor itself, or the hydraulic design of the vessel in the early stage of the reactor conception.

  13. Fresh-Core Reload of the Neutron Radiography (NRAD) Reactor with Uranium(20)-Erbium-Zirconium-Hydride Fuel

    Energy Technology Data Exchange (ETDEWEB)

    John D. Bess; Thomas L. Maddock; Margaret A. Marshall; Leland M. Montierth

    2014-03-01

    The neutron radiography (NRAD) reactor is a 250 kW TRIGA® (Training, Research, Isotopes, General Atomics) Mark II , tank-type research reactor currently located in the basement, below the main hot cell, of the Hot Fuel Examination Facility (HFEF) at the Idaho National Laboratory (INL). It is equipped with two beam tubes with separate radiography stations for the performance of neutron radiography irradiation on small test components. The 60-fuel-element operational core configuration of the NRAD LEU TRIGA reactor has been evaluated as an acceptable benchmark experiment. The initial critical configuration developed during the fuel loading process, which contains only 56 fuel elements, has not been evaluated as it is very similar to the evaluated core configuration. The benchmark eigenvalue is 1.0012 ± 0.0029. Calculated eigenvalues differ significantly (~±1%) from the benchmark eigenvalue and have demonstrated sensitivity to the thermal scattering treatment of hydrogen in the U-Er-Zr-H fuel.

  14. Fresh-Core Reload of the Neutron Radiography (NRAD) Reactor with Uranium(20)-Erbium-Zirconium-Hydride Fuel

    Energy Technology Data Exchange (ETDEWEB)

    John D. Bess; Thomas L. Maddock; Margaret A. Marshall; Leland M. Montierth

    2013-03-01

    The neutron radiography (NRAD) reactor is a 250 kW TRIGA® (Training, Research, Isotopes, General Atomics) Mark II , tank-type research reactor currently located in the basement, below the main hot cell, of the Hot Fuel Examination Facility (HFEF) at the Idaho National Laboratory (INL). It is equipped with two beam tubes with separate radiography stations for the performance of neutron radiography irradiation on small test components. The initial critical configuration developed during the fuel loading process, which contains only 56 fuel elements, has been evaluated as an acceptable benchmark experiment. The 60-fuel-element operational core configuration of the NRAD LEU TRIGA reactor has also been evaluated as an acceptable benchmark experiment. Calculated eigenvalues differ significantly (~±1%) from the benchmark eigenvalue and have demonstrated sensitivity to the thermal scattering treatment of hydrogen in the U-Er-Zr-H fuel.

  15. Fresh-Core Reload of the Neutron Radiography (NRAD) Reactor with Uranium(20)-Erbium-Zirconium-Hydride Fuel

    Energy Technology Data Exchange (ETDEWEB)

    John D. Bess; Thomas L. Maddock; Margaret A. Marshall; Leland M. Montierth

    2011-03-01

    The neutron radiography (NRAD) reactor is a 250 kW TRIGA® (Training, Research, Isotopes, General Atomics) Mark II , tank-type research reactor currently located in the basement, below the main hot cell, of the Hot Fuel Examination Facility (HFEF) at the Idaho National Laboratory (INL). It is equipped with two beam tubes with separate radiography stations for the performance of neutron radiography irradiation on small test components. The 60-fuel-element operational core configuration of the NRAD LEU TRIGA reactor has been evaluated as an acceptable benchmark experiment. The initial critical configuration developed during the fuel loading process, which contains only 56 fuel elements, has not been evaluated as it is very similar to the evaluated core configuration. The benchmark eigenvalue is 1.0012 ± 0.0029. Calculated eigenvalues differ significantly (~±1%) from the benchmark eigenvalue and have demonstrated sensitivity to the thermal scattering treatment of hydrogen in the U-Er-Zr-H fuel.

  16. Reactors

    CERN Document Server

    International Electrotechnical Commission. Geneva

    1988-01-01

    This standard applies to the following types of reactors: shunt reactors, current-limiting reactors including neutral-earthing reactors, damping reactors, tuning (filter) reactors, earthing transformers (neutral couplers), arc-suppression reactors, smoothing reactors, with the exception of the following reactors: small reactors with a rating generally less than 2 kvar single-phase and 10 kvar three-phase, reactors for special purposes such as high-frequency line traps or reactors mounted on rolling stock.

  17. Initial Evaluation of the Heat-Affected Zone, Local Embrittlement Phenomenon as it Applies to Nuclear Reactor Vessels

    Energy Technology Data Exchange (ETDEWEB)

    McCabe, D.E.

    1999-09-01

    The objective of this project was to determine if the local brittle zone (LBZ) problem, encountered in the testing of the heat-affected zone (HAZ) part of welds in offshore platform construction, can also be found in reactor pressure vessel (RPV) welds. Both structures have multipass welds and grain coarsening along the fusion line. Literature was obtained that described the metallurgical evidence and the type of research work performed on offshore structure welds.

  18. Mechanical properties of the as-forged and the forged-and-milled steels for reactor pressure vessel

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Bong Sang; Yoon, Ji Hyun; Kim, Joo Hak; Oh, Yong Jun; Hong, Jun Hwa [Korea Atomic Energy Research Institute, Taejeon (Korea)

    2001-04-01

    The mechanical properties of the as-forged and the forged and milled SA508-Gr.3 reactor pressure vessel steels were evaluated. The full Charpy impact curves obtained for four different locations in test materials. The various data including yield strengths, tensile strengths, elongations were obtained from the tensile strengths, elongations were obtained from the tensile test results for two locations in test materials. The detailed test results were integrated and analysed in this report. 6 refs., 7 figs., 5 tabs. (Author)

  19. Study of impact of the AP1000{sup Registered-Sign} reactor vessel upper internals design on fuel performance

    Energy Technology Data Exchange (ETDEWEB)

    Xu Yiban; Conner, Michael; Yuan Kun; Dzodzo, Milorad B.; Karoutas, Zeses; Beltz, Steven A.; Ray, Sumit; Bissett, Teresa A. [Westinghouse Electric Company, Cranberry Township, PA 16066 (United States); Chieng, Ching-Chang, E-mail: cchieng@ess.nthu.edu.tw [National Tsing Hua University, Hsinchu 30043, Taiwan (China); Kao, Min-Tsung; Wu, Chung-Yun [National Tsing Hua University, Hsinchu 30043, Taiwan (China)

    2012-11-15

    One aspect of the AP1000{sup Registered-Sign} reactor design is the reduction in the number of major components and simplification in manufacturing. One design change relative to current Westinghouse reactors of similar size is the reduction in the number of reactor vessel outlet nozzles/hot legs leaving the upper plenum from three to two. With regard to fuel performance, this design difference creates a different flow field in the AP1000 reactor vessel upper plenum (the region above the core). The flow exiting core and entering the upper plenum must turn 90 Degree-Sign , flow laterally through the upper plenum around support structures, and exit through one of the two outlet nozzles. While the flow in the top of the core is mostly axial, there is some lateral flow component as the core flow reacts to the flow field and pressure distribution in the upper plenum. The pressure distribution in the upper plenum varies laterally depending upon various factors including the proximity to the outlet nozzles. To determine how the lateral flow in the top of the AP1000 core compares to current Westinghouse reactors, a computational fluid dynamics (CFD) model of the flow in the upper portion of the AP1000 reactor vessel including the top region of the core, the upper plenum, the reactor vessel outlet nozzles, and a portion of the hot legs was created. Due to geometric symmetry, the computational domain was reduced to a quarter (from the top view) that includes Vulgar-Fraction-One-Quarter of the top of the core, Vulgar-Fraction-One-Quarter of the upper plenum, and Vulgar-Fraction-One-Half of an outlet nozzle. Results from this model include predicted velocity fields and pressure distributions throughout the model domain. The flow patterns inside and around guide tubes clearly demonstrate the influence of lateral flow due to the presence of the outlet nozzles. From these results, comparisons of AP1000 flow versus current Westinghouse plants were performed. Field performance

  20. Analysis of dpa Rates in the HFIR Reactor Vessel using a Hybrid Monte Carlo/Deterministic Method*

    Directory of Open Access Journals (Sweden)

    Risner J.M.

    2016-01-01

    Full Text Available The Oak Ridge High Flux Isotope Reactor (HFIR, which began full-power operation in 1966, provides one of the highest steady-state neutron flux levels of any research reactor in the world. An ongoing vessel integrity analysis program to assess radiation-induced embrittlement of the HFIR reactor vessel requires the calculation of neutron and gamma displacements per atom (dpa, particularly at locations near the beam tube nozzles, where radiation streaming effects are most pronounced. In this study we apply the Forward-Weighted Consistent Adjoint Driven Importance Sampling (FW-CADIS technique in the ADVANTG code to develop variance reduction parameters for use in the MCNP radiation transport code. We initially evaluated dpa rates for dosimetry capsule locations, regions in the vicinity of the HB-2 beamline, and the vessel beltline region. We then extended the study to provide dpa rate maps using three-dimensional cylindrical mesh tallies that extend from approximately 12 in. below to approximately 12 in. above the height of the core. The mesh tally structures contain over 15,000 mesh cells, providing a detailed spatial map of neutron and photon dpa rates at all locations of interest. Relative errors in the mesh tally cells are typically less than 1%.

  1. Analysis of dpa rates in the HFIR reactor vessel using a hybrid Monte Carlo/deterministic method

    Energy Technology Data Exchange (ETDEWEB)

    Blakeman, Edward [Retired

    2016-01-01

    The Oak Ridge High Flux Isotope Reactor (HFIR), which began full-power operation in 1966, provides one of the highest steady-state neutron flux levels of any research reactor in the world. An ongoing vessel integrity analysis program to assess radiation-induced embrittlement of the HFIR reactor vessel requires the calculation of neutron and gamma displacements per atom (dpa), particularly at locations near the beam tube nozzles, where radiation streaming effects are most pronounced. In this study we apply the Forward-Weighted Consistent Adjoint Driven Importance Sampling (FW-CADIS) technique in the ADVANTG code to develop variance reduction parameters for use in the MCNP radiation transport code. We initially evaluated dpa rates for dosimetry capsule locations, regions in the vicinity of the HB-2 beamline, and the vessel beltline region. We then extended the study to provide dpa rate maps using three-dimensional cylindrical mesh tallies that extend from approximately 12 below to approximately 12 above the axial extent of the core. The mesh tally structures contain over 15,000 mesh cells, providing a detailed spatial map of neutron and photon dpa rates at all locations of interest. Relative errors in the mesh tally cells are typically less than 1%.

  2. Evaluation on radioactive waste disposal amount of Kori Unit 1 reactor vessel considering cutting and packaging methods

    Energy Technology Data Exchange (ETDEWEB)

    Choi, Yu Jong; Lee, Seong Cheol; Kim, Chang Lak [KEPCO International Nuclear Graduate School, Ulsan (Korea, Republic of)

    2016-06-15

    Decommissioning of nuclear power plants has become a big issue in South Korea as some of the nuclear power plants in operation including Kori unit 1 and Wolsung unit 1 are getting old. Recently, Wolsung unit 1 received permission to continue operation while Kori unit 1 will shut down permanently in June 2017. With the consideration of segmentation method and disposal containers, this paper evaluated final disposal amount of radioactive waste generated from decommissioning of the reactor pressure vessel in Kori unit 1 which will be decommissioned as the first in South Korea. The evaluation results indicated that the final disposal amount from the top and bottom heads of the reactor pressure vessel with hemisphere shape decreased as they were cut in smaller more effectively than the cylindrical part of the reactor pressure vessel. It was also investigated that 200 L and 320 L radioactive waste disposal containers used in Kyung-Ju disposal facility had low payload efficiency because of loading weight limitation.

  3. Fracture assessment of the Oskarshamn 1 reactor pressure vessel under cold over-pressurization

    Energy Technology Data Exchange (ETDEWEB)

    Sattari-Far, I. [DNV Technical Consulting AB, Stockholm (Sweden)

    2001-03-01

    The major motivation of this study was to develop a methodology for fracture assessment of surface defects in the 01 reactor pressure vessel under cold loading scenarios, particularly the cold over-pressurization event. According to a previous study, the FENIX project, the cold over-pressurization of the O1 reactor is a limiting loading case, as the ductile/brittle transition temperature (RT{sub NDT}) of certain welds in the O1 beltline region may be over 100 deg C at the-end-of-life condition. The FENIX project gave values of the acceptable and critical crack depth to be equal to the thickness of the cladding layer (about 6 mm) under this load case using the ASME K{sub Ic} reference curve methodology. This study is aimed to develop a methodology to give a more precise fracture assessment of the O1 reactor under cold loading scenarios. Some of the main objectives of this study have been as below: To prepare a material which can simulate the mechanical properties and RT{sub NDT} of the O1 reactor at the end-of-life conditions. To conduct a fracture mechanics test program to cover the essential influencing factors, such as crack geometry (shallow and deep cracks) and loading condition (uniaxial and biaxial) on the cleavage fracture toughness. To perform fracture mechanics analyses to identify a suitable methodology for assessment of the experimental results. To study the responses of engineering fracture assessment methods to the experimental results from the clad specimens. To propose a fracture assessment procedure for determination of the acceptable and critical flaw sizes in the 01 reactor under the cold loading events. A test program consisted of experiments on standard SEN(B) specimens and clad beams, containing surface cracks was conducted during the course of this project. A total of nine clad beams and clad cruciform specimens were tested under uniaxial and biaxial loading. The test material is reactor steel of type A 508 Grade B, which is specially heat

  4. ASSESSMENT OF THE POTENTIAL FOR HYDROGEN GENERATION DURING GROUTING OPERATIONS IN THE R AND P REACTOR VESSELS

    Energy Technology Data Exchange (ETDEWEB)

    Wiersma, B.

    2010-05-24

    The R- and P-reactor buildings were retired from service and are now being prepared for deactivation and decommissioning (D and D). D and D activities consist primarily of immobilizing contaminated components and structures in a grout-like formulation. Aluminum corrodes very rapidly when it comes in contact with the alkaline grout materials and as a result produces hydrogen gas. To address this potential deflagration/explosion hazard, the Materials Science and Technology Directorate (MS and T) of the Savannah River National Laboratory (SRNL) has been requested to review and evaluate existing experimental and analytical studies of this issue to determine if any process constraints on the chemistry of the fill material and the fill operation are necessary. Various options exist for the type of grout material that may be used for D and D of the reactor vessels. The grout formulation options include ceramicrete (pH 6-8), low pH portland cement + silica fume grout (pH 10.4), or Portland cement groupt (pH 12.5). The assessment concluded that either ceramicrete or the silica fume grout may be used to safely grout the P-reactor vessel. The risk of accumulation of a flammable mixture of hydrogen between the grout-air interface and the top of the reactor is very low. Portland cement grout, on the other hand, for the same range of process parameters does not provide a margin of safety against the accumulation of flammable gas in the reactor vessel during grouting operations in the P-reactor vessel. It is recommended that this grout not be utilized for this task. The R-reactor vessel cotnains significantly less aluminum based on current facility process knowledge, surface observations, and drawings. Therefore, a Portland cement grout may be considered for grouting operations as well as the other grout formulations. For example, if the grout fill rate is less than 1 inch/min and the grout temperature is maintained at 70 C or less, the risk of hydrogen accumulation during fill

  5. Safety-related considerations for reactor pressure vessels in consideration of hydrogen flaking

    Energy Technology Data Exchange (ETDEWEB)

    Dugan, S.; Herter, K.H.; Schuler, X.; Silcher, H. [Stuttgart Univ. (Germany). MPA

    2013-07-01

    During non-destructive inspection of the reactor pressure vessels in the Belgian nuclear power plants Doel 3 and Tihange 2, a large number of crack-like indications located in the base metal of the core shells were found. As part of the evaluation of these indications, which were identified as flake-like separations (hydrogen flakes), questions arise as to their cause, possible operational growth and the impact on the continued safe operation of the plant. In addition to the operational load cases, possible accidental and beyond design load cases are also of importance. Within the scope of the ''Research Project Component Safety'' (Forschungsvorhaben Komponentensicherheit - FKS) in the time frame mid-1970s to mid-1990s, numerous R and D activities on the material mechanics behavior and qualification of RPV materials were performed at MPA University of Stuttgart. The objectives of these investigations were focused on material mechanical issues related to the integrity of components and included standard material testing as well as component-like large scale specimen tests. Another major objective was the evaluation of non-destructive testing (NDT) methods with respect to their detection capabilities for such defects which developed during the manufacturing process. The investigations also included a study of the conditions favorable for formation or prevention of hydrogen flaking. In the context of this paper, the results from these R and D activities are presented in view of the current issues and in relation to the integrity concept for German RPVs. Ultrasonic testing (UT) techniques applied during manufacturing and during in-service inspections of German RPVs will also be discussed.

  6. Application of Master Curve fracture toughness for reactor pressure vessel integrity assessment in the USA

    Energy Technology Data Exchange (ETDEWEB)

    Server, William; Rosinski, Stan; Lott, Randy; Kim, Charles; Weakland, Dennis

    2002-08-01

    The Master Curve fracture toughness approach has been used in the USA for better defining the transition temperature fracture toughness of irradiated reactor pressure vessel (RPV) steels for end-of-life (EOL) and EOL extension (EOLE) time periods. The first application was for the Kewaunee plant in which the life-limiting material was a circumferential weld metal. Fracture toughness testing of this weld metal corresponding to EOL and beyond EOLE was used to reassess the PTS screening value, RT{sub PTS}, and to develop new operating pressure-temperature curves. The NRC has approved this application using a shift-based methodology and higher safety margins than those proposed by the utility and its contractors. Beaver Valley Unit 1, a First Energy nuclear plant, has performed similar fracture toughness testing, but none of the testing has been conducted at EOL or EOLE at this time. Therefore, extrapolation of the life-limiting plate data to higher fluences is necessary, and the projections will be checked in the next decade by Master Curve fracture toughness testing of all of the Beaver Valley Unit 1 beltline materials (three plates and three welds) at fluences near or greater than EOLE. A supplemental surveillance capsule has been installed in the sister plant, Beaver Valley Unit 2, which has the capability of achieving a higher lead factor while operating under essentially the same environment. The Beaver Valley Unit 1 evaluation has been submitted to the NRC. This paper reviews the shift-based approach taken for the Beaver Valley Unit 1 RPV and presents the use of the RT{sub T{sub 0}} methodology (which evolved out of the Master Curve testing and endorsed through two ASME Code Cases). The applied margin accounts for uncertainties in the various material parameters. Discussion of a direct measurement of RT{sub T{sub 0}} approach, as originally submitted for the Kewaunee case, is also presented.

  7. Technical aspects of the process of segmentation and packaging of the reactor vessel of Jose Cabrera NPP; Aspectos tecnicos del proceso de segmentacion y embalaje de la vasija del reactor de la central nuclear Jose Cabrera

    Energy Technology Data Exchange (ETDEWEB)

    Valdivieso, J. M.; Garcia Castro, R.

    2015-07-01

    Westinghouse is carrying out the segmentation of the reactor pressure vessel (RPV) within the framework of the Dismantling and Decommissioning Project of the Jose Cabrera NPP. The final concept is based on the comprehensive Westinghouse experience in the field of LWR pressure vessel and internals segmentation, and particularly in previous reactor internals segmentation project for Jose Cabrera NPP. This article shows the development of all the activities included: cutting method selection, preparatory works, cutting activities, waste characterization and packaging activities. (Author)

  8. Radiation Damage Assessment in the Reactor Pressure Vessel of the Integral Inherently Safe Light Water Reactor (I2S-LWR)

    Science.gov (United States)

    Flaspoehler, Timothy; Petrovic, Bojan

    2016-02-01

    One of the major limiting factors to nuclear reactors lifetime is the radiation-induced material damage in the Reactor Pressure Vessel (RPV). While older reactors were designed assuming a 40-year operating lifetime, new reactor designs are expected to have lifetimes up to 100 years. For safe operation, the integrity of the RPV must be ensured against significant material property changes. In this work, typical neutron damage indicators are calculated in the RPV of the I2S-LWR (Integral Inherently Safe LWR) Power Plant, including DPA (displacements per atom) and fast neutron fluence (>1 MeV and >0.1MeV). I2S-LWR is a PWR of integral design, which means that its wider downcomer provides additional shielding to the vessel. However, its higher core power density and longer lifetime may offset this advantage. In order to accurately represent the neutron environment for RPV damage assessment, a detailed model based on the preliminary design specifications of the I2S-LWR was developed to be used in the MAVRIC (Monaco with Automated Variance Reduction using Importance Calculations) sequence of the Scale6.1 code package. MAVRIC uses the CADIS (Consistent Adjoint-Driven Importance Sampling) methodology to bias a fixed-source MC (Monte Carlo) simulation. To establish the upper limit of a bounding envelope, a flat-source distribution was used. For the low limit, a center-peaked source was generated using the KENO-VI criticality sequence assuming uniform fresh fuel core. Results based on the preliminary I2S-LWR model show that DPA rates and fast fluence rates are conservatively 75% lower than in typical PWRs being operated currently in the US.

  9. Radiation Damage Assessment in the Reactor Pressure Vessel of the Integral Inherently Safe Light Water Reactor (I2S-LWR

    Directory of Open Access Journals (Sweden)

    Flaspoehler Timothy

    2016-01-01

    Full Text Available One of the major limiting factors to nuclear reactors lifetime is the radiation-induced material damage in the Reactor Pressure Vessel (RPV. While older reactors were designed assuming a 40-year operating lifetime, new reactor designs are expected to have lifetimes up to 100 years. For safe operation, the integrity of the RPV must be ensured against significant material property changes. In this work, typical neutron damage indicators are calculated in the RPV of the I2S-LWR (Integral Inherently Safe LWR Power Plant, including DPA (displacements per atom and fast neutron fluence (>1 MeV and >0.1MeV. I2S-LWR is a PWR of integral design, which means that its wider downcomer provides additional shielding to the vessel. However, its higher core power density and longer lifetime may offset this advantage. In order to accurately represent the neutron environment for RPV damage assessment, a detailed model based on the preliminary design specifications of the I2S-LWR was developed to be used in the MAVRIC (Monaco with Automated Variance Reduction using Importance Calculations sequence of the Scale6.1 code package. MAVRIC uses the CADIS (Consistent Adjoint-Driven Importance Sampling methodology to bias a fixed-source MC (Monte Carlo simulation. To establish the upper limit of a bounding envelope, a flat-source distribution was used. For the low limit, a center-peaked source was generated using the KENO-VI criticality sequence assuming uniform fresh fuel core. Results based on the preliminary I2S-LWR model show that DPA rates and fast fluence rates are conservatively 75% lower than in typical PWRs being operated currently in the US.

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

    DEFF Research Database (Denmark)

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

    1975-01-01

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

  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. ASSESSMENT OF THE POTENTIAL FOR HYDROGEN GENERATION DURING GROUTING OPERATIONS IN THE R AND P REACTOR VESSELS

    Energy Technology Data Exchange (ETDEWEB)

    Wiersma, B.

    2009-10-29

    The R- and P-reactor buildings were retired from service and are now being prepared for deactivation and decommissioning (D&D). D&D activities will consist primarily of immobilizing contaminated components and structures in a grout-like formulation. Aluminum corrodes very rapidly when it comes in contact with the alkaline grout materials and as a result produces hydrogen gas. To address this potential deflagration/explosion hazard, the Materials Science and Technology Directorate (MS&T) of the Savannah River National Laboratory (SRNL) has been requested to review and evaluate existing experimental and analytical studies of this issue to determine if any process constraints on the chemistry of the fill material and the fill operation are necessary. Various options exist for the type of grout material that may be used for D&D of the reactor vessels. The grout formulation options include ceramicrete (pH 6-8), low pH portland cement + silica fume grout (pH 10.4), or portland cement grout (pH 12.5). The assessment concluded that either ceramicrete or the silica fume grout may be used to safely grout the R- and P- reactor vessels. The risk of accumulation of a flammable mixture of hydrogen between the grout-air interface and the top of the reactor is very low. Conservative calculations estimate that either ceramicrete or the silica fume grout may be used to safely grout the R- and P- reactor vessels. The risk of accumulation of a flammable mixture of hydrogen between the grout-air interface and the top of the reactor is very low. Although these calculations are conservative, there are some measures that may be taken to further minimize the potential for hydrogen evolution. (1) Minimize the temperature of the grout as much as practical. Lower temperatures will mean lower hydrogen generation rates. Grout temperatures less than 100 C should however, still provide an adequate safety margin for the pH 8 and pH 10.4 grout formulations. (2) Minimize the fill rate as much as

  13. Disclosure of the oscillations in kinetics of the reactor pressure vessel steel damage at fast neutron intensity decreasing

    Science.gov (United States)

    Krasikov, E.; Nikolaenko, V.

    2017-01-01

    Fast neutron intensity influence on reactor materials radiation damage is a critically important question in the problem of the correct use of the accelerated irradiation tests data for substantiation of the materials workability in real irradiation conditions that is low neutron intensity. Investigations of the fast neutron intensity (flux) influence on radiation damage and experimental data scattering reveal the existence of non-monotonous sections in kinetics of the reactor pressure vessels (RPV) steel damage. Discovery of the oscillations as indicator of the self-organization processes presence give reasons for new ways searching on reactor pressure vessel (RPV) steel radiation stability increasing and attempt of the self-restoring metal elaboration. Revealing of the wavelike process in the form of non monotonous parts of the kinetics of radiation embrittlement testifies that periodic transformation of the structure take place. This fact actualizes the problem of more precise definition of the RPV materials radiation embrittlement mechanisms and gives reasons for search of the ways to manage the radiation stability (nanostructuring and so on to stimulate the radiation defects annihilation), development of the means for creating of more stableness self recovering smart materials.

  14. The Surface-to-volume Ratio of the Synthesis Reactor Vessel Governing the Low Temperature Crystallization of ZSM-5

    Directory of Open Access Journals (Sweden)

    Ana Hidayati Mukaromah

    2016-12-01

    Full Text Available Zeolite ZSM-5 is one of major catalysts in petroleum and fine-chemical industries. The synthesis of zeolite ZSM-5 is usually carried out at high temperature above 100 °C using the immense amount of organic structure-directing agents (OSDA. It is interesting to note that fine-tuning the initial gel mixture can be used to enhance the typical slow crystallization rate of ZSM-5. Herein, we report the effect of the surface-to-volume ratio of the reactor vessel to the crystallization of ZSM-5 at low temperature. The surface-to-volume ratio of the reactor vessel could influence the heat-transfer during the synthesis which further governed the crystallization of ZSM-5. It was found that the higher the surface-to-volume of the reactor, the more crystalline of the resulting products. The product with the highest crystallinity exhibited a nearly-spherical morphology composed of smaller ZSM-5 crystallites. This phenomenon allowed the presence of inter-crystallite mesopores which is an advantage for the catalytic reaction using bulky molecules.

  15. Validation of the inspections with ultrasound of the welds of the reactor of ITER vacuum vessel; Validacion de las inspecciones con ultrasonidos de las soldaduras de la Vasija de Vacio del reactor del ITER

    Energy Technology Data Exchange (ETDEWEB)

    Garcia, A.; Fernandez, F.; Perez, C.; Sillero, J. A.

    2013-07-01

    The ITER fusion reactor vacuum vessel has thousands of welding austenitic with shapes and different manufacturing processes. The RCC-MR code, which is that applied to the manufacture of the fusion reactor, requires a volumetric test all of them. This test should be mainly by x-rays and welds where it was not possible to use this method, ultrasonic.09-06.

  16. Methodology of Fuel Burn Up Fitting in VVER-1000 Reactor Core by Using New Ex-Vessel Neutron Dosimetry and In-Core Measurements and its Application for Routine Reactor Pressure Vessel Fluence Calculations

    Directory of Open Access Journals (Sweden)

    Borodkin Pavel

    2016-01-01

    Full Text Available Paper describes the new approach of fitting axial fuel burn-up patterns in peripheral fuel assemblies of VVER-1000 type reactors, on the base of ex-core neutron leakage measurements, neutron-physical calculations and in-core SPND measured data. The developed approach uses results of new ex-vessel measurements on different power units through different reactor cycles and their uncertainties to clear the influence of a fitted fuel burn-up profile to the RPV neutron fluence calculations. The new methodology may be recommended to be included in the routine fluence calculations used in RPV lifetime management and may be taken into account during VVER-1000 core burn-up pattern correction.

  17. Methodology of Fuel Burn Up Fitting in VVER-1000 Reactor Core by Using New Ex-Vessel Neutron Dosimetry and In-Core Measurements and its Application for Routine Reactor Pressure Vessel Fluence Calculations

    Science.gov (United States)

    Borodkin, Pavel; Borodkin, Gennady; Khrennikov, Nikolay

    2016-02-01

    Paper describes the new approach of fitting axial fuel burn-up patterns in peripheral fuel assemblies of VVER-1000 type reactors, on the base of ex-core neutron leakage measurements, neutron-physical calculations and in-core SPND measured data. The developed approach uses results of new ex-vessel measurements on different power units through different reactor cycles and their uncertainties to clear the influence of a fitted fuel burn-up profile to the RPV neutron fluence calculations. The new methodology may be recommended to be included in the routine fluence calculations used in RPV lifetime management and may be taken into account during VVER-1000 core burn-up pattern correction.

  18. The deterministic structural integrity assessment of reactor pressure vessels under pressurized thermal shock loading

    Energy Technology Data Exchange (ETDEWEB)

    Chen, Mingya, E-mail: chenmingya@cgnpc.com.cn; Lu, Feng; Wang, Rongshan; Huang, Ping; Liu, Xiangbin; Zhang, Guodong; Xu, Chaoliang

    2015-07-15

    Highlights: • The conservative and non-conservative assumptions in the codes were shown. • The influence of different loads on the SM was given. • The unloading effect of the cladding was studied. • A concentrated reflection of the safety was shown based on 3-D FE analyses. - Abstract: The deterministic structural integrity of a reactor pressure vessel (RPV) related to pressurized thermal shocks (PTSs) has been extensively studied. While the nil-ductility-transition temperature (RT{sub NDT}) parameter is widely used, the influence of fluence and temperature distributions along the thickness of the base metal wall cannot be reflected in the comparative analysis. This paper introduces the method using a structure safety margin (SM) parameter which is based on a comparison between the material toughness (the fracture initiation toughness K{sub IC} or fracture arrest toughness K{sub Ia}) and the stress intensity factor (SIF) along the crack front for the integrity analysis of a RPV subjected to PTS transients. A 3-D finite element model is used to perform fracture mechanics analyses considering both crack initiation assessment and arrest assessment. The results show that the critical part along the crack front is always the clad-base metal interface point (IP) rather than the deepest point (DP) for either crack initiation assessment or crack arrest assessment under the thermal load. It is shown that the requirement in Regulatory Guide 1.154 that ‘axial flaws with depths less than 20% of the wall thickness and all circumferential flaws should be modeled in infinite length’ may be non-conservative. As the assessment result is often poor universal for a given material, crack and transient, caution is recommended in the safety assessment, especially for the IP. The SIF reduces under the thermal or pressure load if the map cracking (MC) effect is considered. Therefore, the assumption in the ASME and RCCM codes that the cladding should be taken into account in

  19. Development of an integrity evaluation system on the basis of cooperative virtual reality environment for reactor pressure vessel

    Energy Technology Data Exchange (ETDEWEB)

    Kim, J.C.; Choi, J.B.; Kim, Y.J. [SAFE Research Center, Sungkyunkwan Univ., Kyunggi-do (Korea); Choi, Y.H.; Park, Y.W. [Korea Inst. of Nuclear Safety, Daijon (Korea); Yoshimura, S. [Inst. of Environmental Studies, The Univ. of Tokyo, Tokyo (Japan)

    2004-07-01

    Since early 1950's, the fracture mechanics has brought significant impact on structural integrity assessment in a wide range of industries such as power, transportation, civil and petrochemical industries, especially in nuclear power plant industries. For the last two decades, significant efforts have been devoted in developing defect assessment procedures, and as a result, various fitness-for-purpose or fitness-for-service codes have been developed. From another aspect, recent advances in IT (information technologies) bring rapid changes in various engineering fields. IT enables people to share information through network and thus provides concurrent working environment without limitations of locations. For this reason, a network system based on internet or intranet bas been appeared in various fields of business. Evaluating the integrity of critical components is one of the most critical issues in the nuclear industry. In order to evaluate the integrity of structures, a complicated and collaborative procedure is required including periodical in-service inspection, fracture mechanics analysis, etc. And thus, experts in different fields have to cooperate to resolve the integrity problem. In this paper, an integrity evaluation system on the basis of cooperative virtual reality environment for reactor pressure vessel which adopts IT into a structural integrity evaluation procedure for reactor pressure vessel is introduced. The proposed system uses virtual reality (VR) technique, virtual network computing (VNC) and knowledge based programs. This system is able to support 3-dimensional virtual reality environment and provide experts to co-operate each other by accessing related data through internet. The proposed system is expected to provide a more efficient integrity evaluation for reactor pressure vessel. (orig.)

  20. Exchange Flow Characteristics in a Tokamak Vacuum Vessel of Fusion Reactor Under the Loss-of-Vacuum Conditions

    Science.gov (United States)

    Takase, Kazuyuki; Kunugi, Tomoaki; Seki, Yasushi

    1997-06-01

    When a Tokamak vacuum vessel of fusion reactor is broken, buoyancy-driven exchange flows will take place through breaches after the inside pressure of the vacuum vessel (VV) becomes equal to the outside pressure. The exchange flow may bring a mixture of activated dusts and tritium from the inside of the VV to the outside through the breaches. Moreover, the exchange flow may remove decay heat from the plasma-facing components. A preliminary LOVA (Loss Of VAcuum event) apparatus was constructed to investigate quantitative heat transfer characteristics of the exchange flows through the breaches under the LOVA conditions. The results of this study, the relationship between Froude numbers and breach locations in the VV was determined and empirical correlations for the average Froude numbers were derived.

  1. Revisiting the reactor pressure vessel for long-time operation; Revisitando la vasija a presion del reactor para largos tiempos de operacion

    Energy Technology Data Exchange (ETDEWEB)

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

    2013-07-01

    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 LONGIFFE 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)

  2. Modeling the Ductile Brittle Fracture Transition in Reactor Pressure Vessel Steels using a Cohesive Zone Model based approach

    Energy Technology Data Exchange (ETDEWEB)

    Pritam Chakraborty; S. Bulent Biner

    2013-10-01

    Fracture properties of Reactor Pressure Vessel (RPV) steels show large variations with changes in temperature and irradiation levels. Brittle behavior is observed at lower temperatures and/or higher irradiation levels whereas ductile mode of failure is predominant at higher temperatures and/or lower irradiation levels. In addition to such temperature and radiation dependent fracture behavior, significant scatter in fracture toughness has also been observed. As a consequence of such variability in fracture behavior, accurate estimates of fracture properties of RPV steels are of utmost importance for safe and reliable operation of reactor pressure vessels. A cohesive zone based approach is being pursued in the present study where an attempt is made to obtain a unified law capturing both stable crack growth (ductile fracture) and unstable failure (cleavage fracture). The parameters of the constitutive model are dependent on both temperature and failure probability. The effect of irradiation has not been considered in the present study. The use of such a cohesive zone based approach would allow the modeling of explicit crack growth at both stable and unstable regimes of fracture. Also it would provide the possibility to incorporate more physical lower length scale models to predict DBT. Such a multi-scale approach would significantly improve the predictive capabilities of the model, which is still largely empirical.

  3. Evaluation of ductile-brittle transition behavior with neutron irradiation in nuclear reactor pressure vessel steels using small punch test

    Energy Technology Data Exchange (ETDEWEB)

    Kim, M. C.; Lee, B. S. [KAERI, Taejon (Korea, Republic of); Oh, Y. J. [Hanbat National Univ., Taejon (Korea, Republic of)

    2003-10-01

    A Small Punch (SP) test was performed to evaluate the ductile-brittle transition temperature before and after neutron irradiation in Reactor Pressure Vessel (RPV) steels produced by different manufacturing (refining) processes. The results were compared to the standard transition temperature shifts from the Charpy test and Master Curve fracture toughness test in accordance with the ASTM standard E1921. The samples were taken from 1/4t location of the vessel thickness and machined into a 10x10x0.5mm dimension. Irradiation of the samples was carried out in the research reactor at KAERI (HANARO) at about 290 .deg. C of the different fluence levels respectively. SP tests were performed in the temperature range of RT to -196 .deg. C using a 2.4mm diameter ball. For the materials before and after irradiation, SP transition temperatures (T{sub sp}), which are determined at the middle of the upper and lower SP energies, showed a linear correlation with the Charpy index temperature, T{sub 41J}. T{sub sp} from the irradiated samples was increased as the fluence level increased and was well within the deviation range of the unirradiated data. The TSP had a correlation with the reference temperature (T{sub 0}) from the master curve method using a pre-cracked Charpy V-notched (PCVN) specimen.

  4. Design and development of in-vessel viewing periscope for ITER (International Thermonuclear Experimental Reactor)

    Energy Technology Data Exchange (ETDEWEB)

    Obara, Kenjiro; Kakudate, Satoshi; Ito, Akira; Shibanuma, Kiyoshi; Tada, Eisuke [Department of Fusion Engineering Research, Naka Fusion Research Establishment, Japan Atomic Energy Research Institute, Naka , Ibaraki (Japan)

    1999-02-01

    An in-vessel viewing system is essential not only to detect and locate damage of components exposed to plasma, but also to monitor and assist in-vessel maintenance operation. In ITER, the in-vessel viewing system must be capable of operating at high temperature (200degC), under intense gamma radiation (30 kGy/h) and high vacuum or 1 bar inert gas. A periscope-type in-vessel viewing system has been chosen as a reference of the ITER in-vessel viewing system due to its wide viewing capability and durability for sever environments. According to the ITER research and development program, a full-scale radiation hard periscope with a length of 15 m has been successfully developed by the Japan Home Team. The performance tests have been shown sufficient capability at high temperature up to 250degC and radiation resistance over 100 MGy. This report describes the design and R and D results of the ITER in-vessel viewing periscope based on the development of 15-m-length radiation hard periscope. (author)

  5. Design and development of in-vessel viewing periscope for ITER (International Thermonuclear Experimental Reactor)

    Energy Technology Data Exchange (ETDEWEB)

    Obara, Kenjiro; Kakudate, Satoshi; Ito, Akira; Shibanuma, Kiyoshi; Tada, Eisuke [Department of Fusion Engineering Research, Naka Fusion Research Establishment, Japan Atomic Energy Research Institute, Naka , Ibaraki (Japan)

    1999-02-01

    An in-vessel viewing system is essential not only to detect and locate damage of components exposed to plasma, but also to monitor and assist in-vessel maintenance operation. In ITER, the in-vessel viewing system must be capable of operating at high temperature (200degC), under intense gamma radiation (30 kGy/h) and high vacuum or 1 bar inert gas. A periscope-type in-vessel viewing system has been chosen as a reference of the ITER in-vessel viewing system due to its wide viewing capability and durability for sever environments. According to the ITER research and development program, a full-scale radiation hard periscope with a length of 15 m has been successfully developed by the Japan Home Team. The performance tests have been shown sufficient capability at high temperature up to 250degC and radiation resistance over 100 MGy. This report describes the design and R and D results of the ITER in-vessel viewing periscope based on the development of 15-m-length radiation hard periscope. (author)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2001-07-01

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

  7. Physics of hydride fueled PWR

    Science.gov (United States)

    Ganda, Francesco

    The first part of the work presents the neutronic results of a detailed and comprehensive study of the feasibility of using hydride fuel in pressurized water reactors (PWR). The primary hydride fuel examined is U-ZrH1.6 having 45w/o uranium: two acceptable design approaches were identified: (1) use of erbium as a burnable poison; (2) replacement of a fraction of the ZrH1.6 by thorium hydride along with addition of some IFBA. The replacement of 25 v/o of ZrH 1.6 by ThH2 along with use of IFBA was identified as the preferred design approach as it gives a slight cycle length gain whereas use of erbium burnable poison results in a cycle length penalty. The feasibility of a single recycling plutonium in PWR in the form of U-PuH2-ZrH1.6 has also been assessed. This fuel was found superior to MOX in terms of the TRU fractional transmutation---53% for U-PuH2-ZrH1.6 versus 29% for MOX---and proliferation resistance. A thorough investigation of physics characteristics of hydride fuels has been performed to understand the reasons of the trends in the reactivity coefficients. The second part of this work assessed the feasibility of multi-recycling plutonium in PWR using hydride fuel. It was found that the fertile-free hydride fuel PuH2-ZrH1.6, enables multi-recycling of Pu in PWR an unlimited number of times. This unique feature of hydride fuels is due to the incorporation of a significant fraction of the hydrogen moderator in the fuel, thereby mitigating the effect of spectrum hardening due to coolant voiding accidents. An equivalent oxide fuel PuO2-ZrO2 was investigated as well and found to enable up to 10 recycles. The feasibility of recycling Pu and all the TRU using hydride fuels were investigated as well. It was found that hydride fuels allow recycling of Pu+Np at least 6 times. If it was desired to recycle all the TRU in PWR using hydrides, the number of possible recycles is limited to 3; the limit is imposed by positive large void reactivity feedback.

  8. Sensitivity coefficients for the stochastic estimation of the radiation damage to the reactor pressure vessel

    Energy Technology Data Exchange (ETDEWEB)

    Alvarez, C.M.; Hernandez Valle, S. [Centro de Investigaciones Tecnologicas, Nucleares y Ambientales, La Habana (Cuba). E-mail: calvarez@ctn.isctn.edu.cu; svalle@ctn.isctn.edu.cu

    2000-07-01

    The construction of the sensitivity matrix in the case of the vessel radiation damage estimation by Monte Carlo techniques poses new problems related to the uncertainties of the obtained responses. In the case of deterministic calculations, the sensitivity coefficient obtention is a straightforward procedure based on the perturbation formalism through the calculation of the adjoint fluxes. In the paper an alternative procedure implementation based on the differential operator method is described with the modifications needed to the used HEXANN-EVALU code for the response estimations in the VVER-440 pressure vessel. (author)

  9. Final Report for the 1st Surveillance Test of the Reactor Pressure Vessel Material (CAPSULE 2) of Ulchin Nuclear Power Plant Unit 3

    Energy Technology Data Exchange (ETDEWEB)

    Chang, Kee Ok; Kim, Byoung Chul; Lee, Sam Lai (and others)

    2006-12-15

    Surveillance testing for reactor vessel materials is performed in order to evaluate the irradiation embrittlement due to neutrons during operation and set the condition of safe operation of nuclear reactor. The 1st surveillance testing was performed completely by Korea Atomic Energy Research Institute at Taejon after the capsule was transported from Ulchin site including its removal from reactor. Fast neutron fluences for capsules were calculated and various testing including mechanical and chemistry analysis were performed in order to evaluate the integrity of Ulchin unit 3 reactor vessel during the operation until life time. The evaluation results are as follows; Fast neutron fluences for capsule 2 is 4.674E 18n/cm{sup 2}. The bias factor, the ratio of calculation/measurement, was 0.920 for the 1st testing and the calculational uncertainty,7.0% satisfied the requirement of USNRC Reg.Guide 1.190, 20%. The best estimated neutron fluence for reactor vessel inside surface was 3.913E 18n/cm{sup 2} based on the end of 6th fuel cycle and it was predicted that the fluences of vessel inside surface at 16 and 32EFPY would reach 9.249E 18 and 1.834E 19n/cm{sup 2} based on the current calculation. The result through this analysis for Ulchin unit 3 showed that there would be no problem for the pressurized thermal shock(PTS) during the operation until design life.

  10. Final Report for the 1st Surveillance Test of the Reactor Pressure Vessel Material (Capsule 2) of Ulchin Nuclear Power Plant Unit 4

    Energy Technology Data Exchange (ETDEWEB)

    Chang, Kee Ok; Kim, Byoung Chul; Lee, Sam Lai (and others)

    2007-04-15

    Surveillance testing for reactor vessel materials is performed in order to evaluate the irradiation embrittlement due to neutrons during operation and set the condition of safe operation of nuclear reactor. The 1st surveillance testing was performed completely by Korea Atomic Energy Research Institute at Daejon after the capsule was transported from Ulchin site including its removal from reactor. Fast neutron fluences for capsules were calculated and various testing including mechanical and chemistry analysis were performed in order to evaluate the integrity of Ulchin Unit 4 reactor vessel during the operation until life time. The evaluation results are as follows; Fast neutron fluences for capsule 2 is 4.306E+18n/cm{sup 2}. The bias factor, the ratio of calculation/measurement, was 0.918 for the 1st testing and the calculational uncertainty,7.0% satisfied the requirement of USNRC Reg.Guide 1.190, 20%. The best estimated neutron fluence for reactor vessel inside surface was 3.615E+18n/cm{sup 2} based on the end of 6th fuel cycle and it was predicted that the fluences of vessel inside surface at 16 and 32EFPY would reach 8.478E+18 and 1.673E+19n/cm{sup 2} based on the current calculation. The result through this analysis for Ulchin Unit 4 showed that there would be no problem for the pressurized thermal shock(PTS) during the operation until design life.

  11. A preliminary assessment of the effects of heat flux distribution and penetration on the creep rupture of a reactor vessel lower head

    Energy Technology Data Exchange (ETDEWEB)

    Chu, T.Y.; Bentz, J.; Simpson, R. [Sandia National Labs., Albuquerque, NM (United States); Witt, R. [Univ. of Wisconsin, Madison, WI (United States)

    1997-02-01

    The objective of the Lower Head Failure (LHF) Experiment Program is to experimentally investigate and characterize the failure of the reactor vessel lower head due to thermal and pressure loads under severe accident conditions. The experiment is performed using 1/5-scale models of a typical PWR pressure vessel. Experiments are performed for various internal pressure and imposed heat flux distributions with and without instrumentation guide tube penetrations. The experimental program is complemented by a modest modeling program based on the application of vessel creep rupture codes developed in the TMI Vessel Investigation Project. The first three experiments under the LHF program investigated the creep rupture of simulated reactor pressure vessels without penetrations. The heat flux distributions for the three experiments are uniform (LHF-1), center-peaked (LHF-2), and side-peaked (LHF-3), respectively. For all the experiments, appreciable vessel deformation was observed to initiate at vessel wall temperatures above 900K and the vessel typically failed at approximately 1000K. The size of failure was always observed to be smaller than the heated region. For experiments with non-uniform heat flux distributions, failure typically occurs in the region of peak temperature. A brief discussion of the effect of penetration is also presented.

  12. Preliminary Assessment of the Impact on Reactor Vessel dpa Rates Due to Installation of a Proposed Low Enriched Uranium (LEU) Core in the High Flux Isotope Reactor (HFIR)

    Energy Technology Data Exchange (ETDEWEB)

    Daily, Charles R. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2015-10-01

    An assessment of the impact on the High Flux Isotope Reactor (HFIR) reactor vessel (RV) displacements-per-atom (dpa) rates due to operations with the proposed low enriched uranium (LEU) core described by Ilas and Primm has been performed and is presented herein. The analyses documented herein support the conclusion that conversion of HFIR to low-enriched uranium (LEU) core operations using the LEU core design of Ilas and Primm will have no negative impact on HFIR RV dpa rates. Since its inception, HFIR has been operated with highly enriched uranium (HEU) cores. As part of an effort sponsored by the National Nuclear Security Administration (NNSA), conversion to LEU cores is being considered for future HFIR operations. The HFIR LEU configurations analyzed are consistent with the LEU core models used by Ilas and Primm and the HEU balance-of-plant models used by Risner and Blakeman in the latest analyses performed to support the HFIR materials surveillance program. The Risner and Blakeman analyses, as well as the studies documented herein, are the first to apply the hybrid transport methods available in the Automated Variance reduction Generator (ADVANTG) code to HFIR RV dpa rate calculations. These calculations have been performed on the Oak Ridge National Laboratory (ORNL) Institutional Cluster (OIC) with version 1.60 of the Monte Carlo N-Particle 5 (MCNP5) computer code.

  13. Numerical Simulation of the Pressure Distribution in the Reactor Vessel Downcomer Region Fluctuated by the Reactor Coolant Pump

    Energy Technology Data Exchange (ETDEWEB)

    Jeong, Dong Hwa; Jung, Byung Ryul; Jang, Ho Cheol; Yune, Seok Jeong; Kim, Eun Kee [KEPCO EnC, Daejeon (Korea, Republic of)

    2015-10-15

    In this study the numerical simulation of the pressure distribution in the downcomer region resulting from the pressure pulsation by the Reactor Coolant Pump (RCP) is performed using the Finite Difference Method (FDM). Simulation is carried out for the cylindrical shaped 2-dimensional model equivalent to the outer surface of the Core Support Barrel (CSB) of APR1400 and a 1/2 model is adopted based on the bilateral symmetry by the inlet nozzle. The fluid temperature is 555 .deg. F and the forcing frequencies are 120Hz, 240Hz, 360Hz and 480Hz. Simulation results of the axial pressure distributions are provided as the Root Mean Square (RMS) values at the five locations of 0°, 45°, 90°, 135° and 180° in the circumferential direction from the inlet nozzle location. In the study, the numerical simulation of pressure distributions in the downcomer region induced by the RCP was performed using FDM and the results were reviewed. The interference of the waves returned from both boundaries in the axial direction and the source of the sinusoidal wave is shown on the inlet nozzle interface pressure point. It seems that the maximum pressures result from the superposition of the waves reflected from the seating surface and the waves newly arrived from the inlet nozzle interface pressure location.

  14. A study on the irradiation embrittlement and recovery characteristics of light water reactor pressure vessel steels

    Energy Technology Data Exchange (ETDEWEB)

    Chi, Se Hwan; Hong, Jun Hwa; Lee, Bong Sang; Oh, Jong Myung; Song, Sook Hyang; Milan, Brumovsky [NRI Czech (Czech Republic)

    1999-03-01

    The neutron irradiation embrittlement phenomenon of light water RPV steels greatly affects the life span for safe operation of a reactor. Reliable evaluation and prediction of the embrittlement of RPV steels, especially of aged reactors, are of importance to the safe operation of a reactor. In addition, the thermal recovery of embrittled RPV has been recognized as an option for life extension. This study aimed to tracer/refine available technologies for embrittlement characterization and prediction, to prepare relevant materials for several domestic RPV steels of the embrittlement and recovery, and to find out possible remedy for steel property betterment. Small specimen test techniques, magnetic measurement techniques, and the Meechan and Brinkmann's recovery curve analysis method were examined/applied as the evaluation techniques. Results revealed a high irradiation sensitivity in YG 3 RPV steel. Further extended study may be urgently needed. Both the small specimen test technique for the direct determination of fracture toughness, and the magnetic measurement technique for embrittlement evaluation appeared to be continued for the technical improvement and data base preparation. Manufacturing process relevant to the heat treatment appeared to be improved in lowering the irradiation sensitivity of the steel. Further study is needed especially in applying the present techniques to the new structural materials under new irradiation environment of advanced reactors. (author)

  15. Safe operation of a batch reactor: Safe storage of organic peroxides in supply vessels

    NARCIS (Netherlands)

    Steensma, M.; Steensma, Metske; Westerterp, K.R.

    1991-01-01

    In this study, we investigated the limits of safe operation for a cooled reactor, operated batchwise. As an example of a single-phase reaction, we studied the decomposition of t-butyl peroxypivalate, a well-known organic peroxide, undergoing self-heating at relatively low temperatures. If sufficient

  16. Safe operation of a batch reactor: safe storage of organic peroxides in supply vessels

    NARCIS (Netherlands)

    Steensma, Metske; Westerterp, K. Roel

    1991-01-01

    In this study, we investigated the limits of safe operation for a cooled reactor, operated batchwise. As an example of a single-phase reaction, we studied the decomposition of t-butyl peroxypivalate, a well-known organic peroxide, undergoing self-heating at relatively low temperatures. If sufficient

  17. 10 CFR Appendix H to Part 50 - Reactor Vessel Material Surveillance Program Requirements

    Science.gov (United States)

    2010-01-01

    ... within the physical constraints of the system, the neutron spectrum, temperature history, and maximum... light water nuclear power reactors which result from exposure of these materials to neutron irradiation and the thermal environment. Under the program, fracture toughness test data are obtained...

  18. Hydride formation on deformation twin in zirconium alloy

    Science.gov (United States)

    Kim, Ju-Seong; Kim, Sung-Dae; Yoon, Jonghun

    2016-12-01

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

  19. Validation of 3D Code KATRIN For Fast Neutron Fluence Calculation of VVER-1000 Reactor Pressure Vessel by Ex-Vessel Measurements and Surveillance Specimens Results

    Science.gov (United States)

    Dzhalandinov, A.; Tsofin, V.; Kochkin, V.; Panferov, P.; Timofeev, A.; Reshetnikov, A.; Makhotin, D.; Erak, D.; Voloschenko, A.

    2016-02-01

    Usually the synthesis of two-dimensional and one-dimensional discrete ordinate calculations is used to evaluate neutron fluence on VVER-1000 reactor pressure vessel (RPV) for prognosis of radiation embrittlement. But there are some cases when this approach is not applicable. For example the latest projects of VVER-1000 have upgraded surveillance program. Containers with surveillance specimens are located on the inner surface of RPV with fast neutron flux maximum. Therefore, the synthesis approach is not suitable enough for calculation of local disturbance of neutron field in RPV inner surface behind the surveillance specimens because of their complicated and heterogeneous structure. In some cases the VVER-1000 core loading consists of fuel assemblies with different fuel height and the applicability of synthesis approach is also ambiguous for these fuel cycles. Also, the synthesis approach is not enough correct for the neutron fluence estimation at the RPV area above core top. Because of these reasons only the 3D neutron transport codes seem to be satisfactory for calculation of neutron fluence on the VVER-1000 RPV. The direct 3D calculations are also recommended by modern regulations.

  20. Validation of 3D Code KATRIN For Fast Neutron Fluence Calculation of VVER-1000 Reactor Pressure Vessel by Ex-Vessel Measurements and Surveillance Specimens Results

    Directory of Open Access Journals (Sweden)

    Dzhalandinov A.

    2016-01-01

    Full Text Available Usually the synthesis of two-dimensional and one-dimensional discrete ordinate calculations is used to evaluate neutron fluence on VVER-1000 reactor pressure vessel (RPV for prognosis of radiation embrittlement. But there are some cases when this approach is not applicable. For example the latest projects of VVER-1000 have upgraded surveillance program. Containers with surveillance specimens are located on the inner surface of RPV with fast neutron flux maximum. Therefore, the synthesis approach is not suitable enough for calculation of local disturbance of neutron field in RPV inner surface behind the surveillance specimens because of their complicated and heterogeneous structure. In some cases the VVER-1000 core loading consists of fuel assemblies with different fuel height and the applicability of synthesis approach is also ambiguous for these fuel cycles. Also, the synthesis approach is not enough correct for the neutron fluence estimation at the RPV area above core top. Because of these reasons only the 3D neutron transport codes seem to be satisfactory for calculation of neutron fluence on the VVER-1000 RPV. The direct 3D calculations are also recommended by modern regulations.

  1. Design analysis of the molten core confinement within the reactor vessel in the case of severe accidents at nuclear power plants equipped with a reactor of the VVER type

    Science.gov (United States)

    Zvonaryov, Yu. A.; Budaev, M. A.; Volchek, A. M.; Gorbaev, V. A.; Zagryazkin, V. N.; Kiselyov, N. P.; Kobzar', V. L.; Konobeev, A. V.; Tsurikov, D. F.

    2013-12-01

    The present paper reports the results of the preliminary design estimate of the behavior of the core melt in vessels of reactors of the VVER-600 and VVER-1300 types (a standard optimized and informative nuclear power unit based on VVER technology—VVER TOI) in the case of beyond-design-basis severe accidents. The basic processes determining the state of the core melt in the reactor vessel are analyzed. The concept of molten core confinement within the vessel based on the idea of outside cooling is discussed. Basic assumptions and models, as well as the results of calculation of the interaction between molten materials of the core and the wall of the reactor vessel performed by means of the SOCRAT severe accident code, are presented and discussed. On the basis of the data obtained, the requirements on the operation of the safety systems are determined, upon the fulfillment of which there will appear potential prerequisites for implementing the concept of the confinement of the core melt within the reactor in cases of severe accidents at nuclear power plants equipped with VVER reactors.

  2. Wear plates control rod guide tubes top internal reactor vessel C. N. VANDELLOS II; Desgaste placas tubos guia barras de control interno superior vasija del reactor C.N. Vandellos II

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2010-07-01

    The guide tubes for control rods forming part of the upper internals of the reactor vessel, its function is to guide the control rod to permit its insertion in the reactor core. These guide tubes are suspended from the upper support plate which are fixed by bolts and extending to the upper core plate which is fastened by clamping bolts (split pin) to prevent lateral displacement of the guide tubes, while allowing axial expansion.

  3. Influence of structural parameters on the tendency of VVER-1000 reactor pressure vessel steel to temper embrittlement

    Science.gov (United States)

    Gurovich, B.; Kuleshova, E.; Zabusov, O.; Fedotova, S.; Frolov, A.; Saltykov, M.; Maltsev, D.

    2013-04-01

    In this paper the influence of structural parameters on the tendency of steels to reversible temper embrittlement was studied for assessment of performance properties of reactor pressure vessel steels with extended service life. It is shown that the growth of prior austenite grain size leads to an increase of the critical embrittlement temperature in the initial state. An embrittlement heat treatment at the temperature of maximum manifestation of temper embrittlement (480 °C) shifts critical embrittlement temperature to higher values due to the increase of the phosphorus concentration on grain boundaries. There is a correlation between phosphorus concentration on boundaries of primary austenite grains and the share of brittle intergranular fracture (that, in turn, depends on impact test temperature) in the fracture surfaces of the tested Charpy specimens.

  4. Status and Perspectives of Nuclear Reactor Pressure Vessel Life Extension up to 60 Years Operation in Belgium

    Energy Technology Data Exchange (ETDEWEB)

    Lucon, E.; Chaouadi, R.; Scibetta, M.; Van Walle, E.

    2009-09-15

    The scope of this report involves a safety evaluation of the reactor pressure vessel (RPV) against neutron embrittlement, in the most severely irradiation region (belt line) and in the event of a pressurized thermal shock. The irreplaceable RPV is considered to be the most critical component for lifetime considerations of the nuclear power plant. However, an application for operation extension will also depend upon a number of additional considerations, including the technical assessment of other plant components, as well as non-technical arguments (e.g. political, environmental, economical, strategical that are outside the scope this report. In the hypothesis of a request for operation extension, it is the responsibility of the utilities to provide the safety authorities with an exhaustive dossier demonstrating that safe extended operation is guaranteed. The role of the safety authorities is to critically evaluate the safety dossier for eventually granting the operation extension.

  5. Structural evaluation of the Shippingport Reactor Pressure Vessel and Neutron Shield Tank package for impact and puncture loads

    Energy Technology Data Exchange (ETDEWEB)

    Fischer, L.E.; Chou, C.K.; Lo, T.; Schwartz, M.W.

    1988-06-01

    A structural evaluation of Shippingport Reactor Pressure Vessel and Neutron Shield Tank package for impact and puncture loads under the normal and hypothetical accident conditions of 10 CFR 71 was performed. Component performance criteria for the Shippingport package and the corresponding structural acceptance criteria for these components were developed based on a review of the package geometry, the planned transport environment, and the external radiation standards and dispersal limits of 10 CFR 71. The evaluation was performed using structural analysis methods. A demonstration combining simplified model tests and nonlinear finite element analyses was made to substantiate the structural analysis methods used to evaluate the Shippingport package. The package was analyzed and the results indicate that the package meets external radiation standards and release limits of 10 CFR 71. 13 refs., 50 figs., 19 tabs.

  6. Standard Guide for Predicting Radiation-Induced Transition Temperature Shift in Reactor Vessel Materials, E706 (IIF)

    CERN Document Server

    American Society for Testing and Materials. Philadelphia

    2002-01-01

    1.1 This guide presents a method for predicting reference transition temperature adjustments for irradiated light-water cooled power reactor pressure vessel materials based on Charpy V-notch 30-ftlbf (41-J) data. Radiation damage calculative procedures have been developed from a statistical analysis of an irradiated material database that was available as of May 2000. The embrittlement correlation used in this guide was developed using the following variables: copper and nickel contents, irradiation temperature, and neutron fluence. The form of the model was based on current understanding for two mechanisms of embrittlement: stable matrix damage (SMD) and copper-rich precipitation (CRP); saturation of copper effects (for different weld materials) was included. This guide is applicable for the following specific materials, copper, nickel, and phosphorus contents, range of irradiation temperature, and neutron fluence based on the overall database: 1.1.1 MaterialsA 533 Type B Class 1 and 2, A302 Grade B, A302 G...

  7. Innovative inspection system for reactor pressure vessels; Innovative Pruefsysteme fuer Reaktordruckbehaelter

    Energy Technology Data Exchange (ETDEWEB)

    Mertens, K.; Trautmann, H.

    1999-08-01

    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) [Deutsch] Die vorgestellten kleinen, flexiblen und modernen Schwimmsysteme (DELPHIN-Manipulatoren und MIDAS-U-Boote) sind innovative Systeme fuer die Reduzierung der Aufwaende und Zeit zur Pruefung des Reaktordruckbehaelters und damit zur Reduktion der Revisionszeiten der Reaktoranlagen. (orig.)

  8. Boron Hydrides

    Science.gov (United States)

    1946-07-01

    of direct interest could be b.P.4d. ’Thus the discovory of a now proj.ect, since silano is probably too readily infla-zmablo for practical usc’ this...devoted, ho specc4fie compounds vhitih a’-ould be tocdte at prescnt arc: nron tiy * silano , %;2.SiFi3 , diothyl sila~no, (C2 115 )2 Si112, mono r.-rop; ! (n...Bcrohydrido or Li h.... I .A-4A- The prepuation of Silano med of Stannane by the interaction or lithium aluzirun hydride v-ithl silicon tetrtchiorido and

  9. Ex-vessel coolability and energetics of steam explosions in nordic light water reactors

    Energy Technology Data Exchange (ETDEWEB)

    Park, H.S.; Dinh, T.N. [Royal Institute of Technology (Sweden)

    2007-04-15

    The report summarizes activities conducted at the Division of Nuclear Power Safety, Royal Institute of Technology-Sweden (KTH-NPS) within the ExCoolSe project during the year 2005, which is a transition year for the KTH-NPS program. The ExCoolSe project supported by NKS contributes to the severe accident research at KTH-NPS concurrently supported by APRI, HSK and EU SARNET. The main objective in ExCoolSe project is to scrutinize research on risk-significant safety issues related to severe accident management (SAM) strategy adopted for Nordic BWR plants, namely the Ex-vessel Coolability and Energetic Steam explosion. The work aims to pave way toward building a tangible research framework to tackle these long-standing safety issues. Chapter 1 describes the project objectives and work description. Chapter 2 provides a critical assessment of research results obtained from several past programs at KTH. This includes review of key data, insights and implications from POMECO (Porous Media Coolability) program, COMECO (Corium Melt Coolability) program, SIMECO (Study of In-Vessel Melt Coolability) program, and MISTEE (Micro-Interactions in Steam Explosion Experiments) program. Chapter 3 discusses the rationale of the new research program focusing on the SAM issue resolution. The program emphasizes identification and qualification of physics-based limiting mechanisms for both in-vessel phenomena (melt progression and debris coolability in the lower head, vessel failure), and ex-vessel phenomena. Chapter 4 introduces research results from the newly established DEFOR (Debris Formation) program and the ongoing MISTEE program. The focus of DEFOR is fulfill an apparent gap in the contemporary knowledge of severe accidents, namely mechanisms which govern the debris bed formation and bed characteristics. The later control the debris bed coolability. In the MISTEE program, methods for image synchronization and data processing were developed and tested, which enable processing of

  10. Methodological developments in the field of structural integrity analyses of large scale reactor pressure vessels in Hungary

    Directory of Open Access Journals (Sweden)

    Tamás Fekete

    2016-03-01

    Full Text Available Buildings, structures and systems of large scale and high value (e.g. conventional and nuclear power plants, etc. are designed for a certain, limited service lifetime. If the standards and guidelines of the time are taken into account during the design process, the resulting structures will operate safely in most cases. However, in the course of technical history there were examples of unusual, catastrophic failures of structures, even resulting in human casualties. Although the concept of Structural Integrity first appeared in industrial applications only two-three decades ago, its pertinence has been growing higher ever since. Four nuclear power generation units have been constructed in Hungary, more than 30 years ago. In every unit, VVER-440 V213 type light-water cooled, light-water moderated, pressurized water reactors are in operation. Since the mid-1980s, Pressurized Thermal Shock (PTS analyses of Reactor Pressure Vessels (RPV have been conducted in Hungary, where the concept of structural integrity was the basis of research and development. In the first part of the paper, a short historic overview is given, where the origins of the Structural Integrity concept are presented, and the beginnings of Structural Integrity in Hungary are summarized. In the second part, a new conceptual model of Structural Integrity is introduced. In the third part, a brief description of the VVER-440 V213 type RPV and its surrounding primary system is presented. In the fourth part, a conceptual model developed for PTS Structural Integrity Analyses is explained.

  11. An analysis of the hydrogen bubble concerns in the three-mile island unit-2 reactor vessel

    Science.gov (United States)

    Gordon, S.; Schmidt, K. H.; Honekamp, J. R.

    On 30 March 1979, two days after the accident at the Three-Mile Island Reactor near Harrisburg, Pennsylvania, press reports appeared about a non-condensable bubble in the reactor vessel. This bubble was said to consist mainly of hydrogen, and to grow rapidly, possibly due to the development of oxygen. Danger of explosion was reported to be imminent. We analyzed all possible sources of non-condensable gases, including radiolysis, and determined that a continuing growth of the bubble during several days after the accident was not possible. Our main conclusions were the following: (1) Most of the initial hydrogen in the bubble was produced by the reaction of the Zircalloy cladding with the super-heated water. (2) During the first 16 hr after shutdown, when boiling of the primary coolant water took place, in the worst case stoichiometric amounts of hydrogen and oxygen could have been produced by radiolysis, leading to a maximum amount of oxygen in the bubble, of 0.7% of the hydrogen, which is well below the explosion limit. (3) After this 16 hr period, when boiling had totally ceased, no further oxygen could have been produced by radiolysis of the primary cooling water. On the contrary, oxygen was recombined with hydrogen due to radiolysis at such a rate that the oxygen in the water was completely removed in less than five minutes. The subsequent rate of removal of oxygen from the bubble by dissolution and radiolysis depended essentially on the rate of dissolution.

  12. Probabilistic Fracture Mechanics Analysis of Boling Water Reactor Vessel for Cool-Down and Low Temperature Over-Pressurization Transients

    Directory of Open Access Journals (Sweden)

    Jeong Soon Park

    2016-04-01

    Full Text Available The failure probabilities of the reactor pressure vessel (RPV for low temperature over-pressurization (LTOP and cool-down transients are calculated in this study. For the cool-down transient, a pressure–temperature limit curve is generated in accordance with Section XI, Appendix G of the American Society of Mechanical Engineers (ASME code, from which safety margin factors are deliberately removed for the probabilistic fracture mechanics analysis. Then, sensitivity analyses are conducted to understand the effects of some input parameters. For the LTOP transient, the failure of the RPV mostly occurs during the period of the abrupt pressure rise. For the cool-down transient, the decrease of the fracture toughness with temperature and time plays a main role in RPV failure at the end of the cool-down process. As expected, the failure probability increases with increasing fluence, Cu and Ni contents, and initial reference temperature-nil ductility transition (RTNDT. The effect of warm prestressing on the vessel failure probability for LTOP is not significant because most of the failures happen before the stress intensity factor reaches the peak value while its effect reduces the failure probability by more than one order of magnitude for the cool-down transient.

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

    CERN Document Server

    Puls, Manfred P

    2012-01-01

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

  14. Status of Knowledge of Radiation Embrittlement in USA Reactor Pressure Vessel Steels.

    Science.gov (United States)

    1982-02-01

    NRC-FIN-5528 UNCLASSIFIED NRL-R-,737 NUREG -CR-2511 Ni." EEEIIEIIIEI mEE/hhhE AD ~ NUREG /CR-251 1 0 AD A I INRL Memo Rpt 4737 Status of Knowledge of...J2.75 ti a Tec nica for ati Irv Ce Spr ~~~ Sid 1 NUREG /CR-2511 NRL Memo Rpt 4737 R5 Status of Knowledge of Radiation Embrittlement in USA Reactor...vendor reports and correspondence; Commission papers; and applicant and licensee documents and correspondence. The following documents in the NUREG

  15. Effects of Geometry of Reactor Pressure Vessel Upper Head Control Rod Drive Mechanism Penetration Nozzles on J-Groove Weld Residual Stress

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Ju Hee; Kim, Yun Jae; Bae, Hong Yeol [Korea University, Seoul (Korea, Republic of); and others

    2011-10-15

    In pressurized water reactors (PWRs), the reactor pressure vessel (RPV) upper head contains numerous control rod drive mechanism (CRDM) nozzles. In the last 10 years, the incidences of cracking in alloy 600 CRDM nozzles and their associated welds has increased significantly. Several axial and circumferential cracks have been found in CRDM nozzles in European PWRs and U.S. nuclear power plants. These cracks are caused by primary water stress corrosion cracking (PWSCC) and have been shown to be driven by welding residual stresses and operational stresses in the weld region. Therefore, detailed finite-element (FE) simulations for the Korea Nuclear Reactor Pressure Vessel have been conducted in order to predict the magnitudes of the weld residual stresses in the tube materials. In particular, the weld residual stress results are compared in terms for nozzle location, geometry factor ro/t, geometry of fillet, and adjacent nozzle.

  16. 60th Anniversary of electricity production from light water reactors: Historical review of the contribution of materials science to the safety of the pressure vessel

    Science.gov (United States)

    van Duysen, J. C.; Meric de Bellefon, G.

    2017-02-01

    The first light water nuclear reactor dedicated to electricity production was commissioned in Shippingport, Pennsylvania in the United States in 1957. Sixty years after the event, it is clear that this type of reactor will be a major source of electricity and one of the key solutions to limit climate change in the 21st century. This article pays homage to the teams that contributed to this achievement by their involvement in research and development and their determination to push back the frontiers of knowledge. Via a few examples of scientific or technological milestones, it describes the evolution of ideas, models, and techniques during the last 60 years, and gives the current state-of-the-art in areas related to the safety of the reactor pressure vessel. Among other topics, it focuses on vessel manufacturing, steel fracture mechanics analysis, and understanding of irradiation-induced damage.

  17. Numerical simulation and performance test of metal hydride hydrogen storage system

    Directory of Open Access Journals (Sweden)

    Tzu-Hsiang Yen, Bin-Hao Chen, Bao-Dong Chen

    2011-05-01

    Full Text Available Metal hydride reactors are widely used in many industrial applications, such as hydrogen storage, thermal compression, heat pump, etc. According to the research requirement of metal hydride hydrogen storage, the thermal analyses have been implemented in the paper. The metal hydride reaction beds are considered as coupled cylindrical tube modules which combine the chemical absorption and desorption in metal hydride. The model is then used metal hydride LaNi5 as an example to predict the performance of metal hydride hydrogen storage devices, such as the position of hydration front and the thermal flux. Under the different boundary condition the characteristics of heat transfer and mass transfer in metal hydride have influence on the hydrogen absorption and desorption. The researches revealed that the scroll design can improve the temperature distribution in the reactor and the porous tube for directing hydrogen can increase the penetration depth of hydride reaction to decrease the hydrogen absorption time.

  18. Component test for validation of the design life of reactor vessel wall of the fast breeder reactor SNR 300 regarding creep-fatigue. Modellversuch zur Absicherung der Auslegung der Reaktortankwand des SNR 300 hinsichtlich Kriechermuedung

    Energy Technology Data Exchange (ETDEWEB)

    Lohse, U. (Interatom GmbH, Bergisch Gladbach (Germany)); Laue, H. (Interatom GmbH, Bergisch Gladbach (Germany)); Rathjen, P. (Interatom GmbH, Bergisch Gladbach (Germany)); Maile, K. (Staatliche Materialpruefungsanstalt, Stuttgart (Germany)); Eckert, W. (Staatliche Materialpruefungsanstalt, Stuttgart (Germany)); Purper, H. (Staatliche Materialpruefungsanstalt, Stuttgart (Germany))

    1991-01-01

    The design of the reactor vessel wall is tested under a long-term creep-fatigue stress, with the aid of a similar components test. For this purpose, the results of calculation are compared with the experimental results concerning deformation depending on point and time, and damage depending on the initial state of material, point and time. (orig./HP)

  19. Low cycle fatigue behavior of a ferritic reactor pressure vessel steel

    Energy Technology Data Exchange (ETDEWEB)

    Sarkar, Apu, E-mail: asarkar@barc.gov.in; Kumawat, Bhupendra K.; Chakravartty, J.K.

    2015-07-15

    The cyclic stress–strain response and the low cycle fatigue (LCF) behavior of 20MnMoNi55 pressure vessel steel were studied. Tensile strength and LCF properties were examined at room temperature (RT) using specimens cut from rolling direction of a rolled block. The fully reversed strain-controlled LCF tests were conducted at a constant total strain rate with different axial strain amplitude levels. The cyclic strain–stress relationships and the strain–life relationships were obtained through the test results, and related LCF parameters of the steel were calculated. The studied steel exhibits cyclic softening behavior. Furthermore, analysis of stabilized hysteresis loops showed that the steel exhibits non-Masing behavior. Complementary scanning electron microscopy examinations were also carried out on fracture surfaces to reveal dominant damage mechanisms during crack initiation, propagation and fracture. Multiple crack initiation sites were observed on the fracture surface. The investigated LCF behavior can provide reference for pressure vessel life assessment and fracture mechanisms analysis.

  20. Low cycle fatigue behavior of a ferritic reactor pressure vessel steel

    Science.gov (United States)

    Sarkar, Apu; Kumawat, Bhupendra K.; Chakravartty, J. K.

    2015-07-01

    The cyclic stress-strain response and the low cycle fatigue (LCF) behavior of 20MnMoNi55 pressure vessel steel were studied. Tensile strength and LCF properties were examined at room temperature (RT) using specimens cut from rolling direction of a rolled block. The fully reversed strain-controlled LCF tests were conducted at a constant total strain rate with different axial strain amplitude levels. The cyclic strain-stress relationships and the strain-life relationships were obtained through the test results, and related LCF parameters of the steel were calculated. The studied steel exhibits cyclic softening behavior. Furthermore, analysis of stabilized hysteresis loops showed that the steel exhibits non-Masing behavior. Complementary scanning electron microscopy examinations were also carried out on fracture surfaces to reveal dominant damage mechanisms during crack initiation, propagation and fracture. Multiple crack initiation sites were observed on the fracture surface. The investigated LCF behavior can provide reference for pressure vessel life assessment and fracture mechanisms analysis.

  1. Specialist meeting on leak before break in reactor piping and vessels

    Energy Technology Data Exchange (ETDEWEB)

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

    1997-04-01

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

  2. Application of advanced master curve approaches on WWER-440 reactor pressure vessel steels

    Energy Technology Data Exchange (ETDEWEB)

    Viehrig, Hans-Werner [Forschungszentrum Rossendorf e.V. (Germany)]. E-mail: h.w.viehrig@fz-rossendorf.de; Scibetta, Marc [SCK-CEN, Reactor Materials Research (Belgium); Wallin, Kim [VTT Industrial Systems, Materials and Structural Integrity (Finland)

    2006-08-15

    The master curve (MC) approach used to measure the transition temperature, T , was standarised in the ASTM Standard Test Method E 1921 in 1997. The basic MC approach for analysis of fracture test results is intended for macroscopically homogeneous steels with a body centred cubic (ferritic) structure only. In reality, due to the manufacturing process, the steels in question are seldom fully macroscopically homogeneous. The fracture toughness values measured on Charpy size SE(B) specimens of base metal from the Greifswald Unit 8 rector pressure vessel (RPV) show large scatter. The basic MC evaluation following ASTM E1921 supplies a MC with many fracture toughness values which lie below the 5% fracture probability line. It is therefore suspected that this material is macroscopically inhomogeneous. In this paper, two recent extensions of the MC for inhomogeneous materials are applied to these fracture toughness data.

  3. 10 years of control, repair and replacement of reactor vessel heads; Dix ans de controles, reparations et remplacements de couvercles de cuves

    Energy Technology Data Exchange (ETDEWEB)

    Goyau, G. [AREVA/FRAMATOME ANP, Div. Service Composants Primaires, 75 - Paris (France)

    2004-06-01

    The occurring of a leak on the primary cooling system at the level of the vessel head of the Bugey-3 reactor during a prescribed hydrostatic test in 1991 has led to more than 10 years of controls, reparations and replacements of vessel heads on French or exported units. This article recalls and details the consequences of this generic defects concerning: -) the devising of special control equipment based on Foucault currents and ultra-sounds, -) the designing of specific repairing tools, and -) the strategy of replacing the vessel head that is now followed in most cases. By end 2003 the total number of vessel head that Framatome-ANP had changed throughout the world, reached 53 units. (A.C.)

  4. Evaluation of neutron irradiation embrittlement in the Korean reactor pressure vessel steels(I) (1st progress report)

    Energy Technology Data Exchange (ETDEWEB)

    Hong, Jun Hwa; Lee, Bong Sang; Park, Duck Gun; Byun, Tak Sang; Kim, Joo Hag; Oh, Yong Jun; Yoon, Ji Hyun; Chi, Sei Hwan; Kuk, Il Hyun [Korea Atomic Energy Research Institute, Taejon (Korea)

    1998-10-01

    The SA508-3 reactor pressure vessel materials degrade due to the application at high temperature, high pressure, and neutron irradiation. In the present study it is planned to examine the effects of neutron irradiation on the properties for assessing the integrity of domestic reactors. The key tests are the Charpy impact test, tensile test, static and dynamic fracture toughness test, J-R test. The additional tests for obtaining basic material properties, such as micro-hardness, microstructural properties, small punch energy etc., are also performed. The irradiation tests are being performed at HANARO of KAERI through the instrumented capsules designed by KAERI and the post-irradiation tests are being performed at IMEF(Irradiated Material Evaluation Facility) of material (UCN-4), Si+Al (YGN-5), UCN-4 weld metal, and UCN-4 HAZ. In the irradiation test the temperature should be controlled in the range of 290 {+-} 10 deg C and the test materials would be irradiated to 2 to 3 neutron fluence levels including the end-of-life fluence. The status of performing this project is that (1) the key data on mechanical properties, mainly related to the fracture toughness, of the unirradiated materials have been obtained, (2) the irradiation of the 1st instrumented capsule, a preliminary test capsule containing miniature specimens, has been completed and is being stored for testing in IMEF, and (3) the 2nd instrumented capsule is being manufactured and will be irradiated in the beginning or 1999. This report includes mainly the experimental methods and results. The status of the design and manufacturing of the instrumented capsules and specimens was also briefly described. (author). 13 refs., 15 figs., 10 tabs.

  5. The 4th surveillance test and evaluation of the reactor pressure vessel material (capsule W) of Younggwang nuclear power plant unit1

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Byoung Chul; Lee, Sam Lai; Choi, Kwon Jae [Korea Atomic Energy Research Institute, Taejeon (Korea)

    2001-08-01

    Surveillance testing for reactor vessel materials is performed in order to evaluate the irradiation embrittlement due to neutrons during operation and set the condition of safe operation of nuclear reactor. The 4th surveillance testing was performed completely by Korea Atomic Energy Research Institute at Taejeon after the capsule was transported from Yonggwang site including its removal from reactor. Fast neutron fluences for capsules were calculated and various testing including mechanical and chemistry analysis were performed in order to evaluate the integrity of Yonggwang unit 1 reactor vessel during the operation until life time. The evaluation results are as follows; Fast neutron fluences for capsules U, V, X and W are 5.555E+18, 1.662E+19, 3.358E+19, and 4.521E+19 n/cm{sup 2}, respectively. The bias factor, the ratio of measurement versus calculation, was 0.859 for the 1st through 4th testing and the calculational uncertainty, 11.80% satisfied the requirement of USNRC Reg.Guide 1.190, 20%. The best estimated neutron fluence for reactor vessel inside surface was 1.551E+19 n/cm{sup 2} based on the end of 12th fuel cycle and it was predicted that the fluences of vessel inside surface at 32, 40, 48 and 56EFPY would reach 3.929E+19, 4.880E+19, 5.831E+19 and 6.782E+19 n/cm{sup 2} based on the current calculation. The result through this analysis for Yonggwang unit 1 showed that there would be no problem for the pressurized thermal shock(PTS) during the operation until design life. 4 refs., 41 figs., 35 tabs. (Author)

  6. Final report for the 5th surveillance test of the reactor pressure vessel material (capsule Y) of Yonggwang Nuclear Power Plant unit 2

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Sam Lai; Kim, ByoungChul; Chang, Kee Ok (and others)

    2006-02-15

    Surveillance testing for reactor vessel materials is performed in order to evaluate the irradiation embrittlement due to neutrons during operation and set the condition of safe operation of nuclear reactor. The 5th surveillance testing was performed completely by Korea Atomic Energy Research Institute at Daejeon after the capsule was transported from Yonggwang site including its removal from reactor. Fast neutron fluences for capsules were calculated and various testing including mechanical and chemistry analysis were performed in order to evaluate the integrity of Yonggwang unit 2 reactor vessel during the operation until life time. The evaluation results are as follows; Fast neutron fluences for capsules U, V, X, W and Y are 5.777E+18, 1.5371E+19, 3.7634E+19, 4.3045E+19, and 4.8662E+19n/cm{sup 2}, respectively. The bias factor, the ratio of calculation/measurement, was 0.953 for the 1st through 5th testing and the calculational uncertainty,7.2% satisfied the requirement of USNRC Reg.Guide 1.190, 20%. The best estimated neutron fluence for reactor vessel inside surface was 1.659E+19n/cm{sup 2} based on the end of 13th fuel cycle and it was predicted that the fluences of vessel inside surface at 32, 48, 56 and 64EFPY would reach 3.625E+19, 5.293E+19, 6.127E+19 and 6.960E+19n/cm{sup 2} based on the current calculation. The result through this analysis for Yonggwang unit 2 showed that there would be no problem for the Pressurized Thermal Shock(PTS) during the operation until design life.

  7. The 4th surveillance test and evaluation of the reactor pressure vessel material (capsule W) of Yonggwang nuclear power plant unit 2

    Energy Technology Data Exchange (ETDEWEB)

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

    2001-02-01

    Surveillance testing for reactor vessel materials is performed in order to evaluate the irradiation embrittlement due to neutrons during operation and set the condition of safe operation of nuclear reactor. The 4th surveillance testing was performed completely by Korea Atomic Energy Research Institute at Taejeon after the capsule was transported from Kori site including its removal from reactor. Fast neutron fluences for capsules were calculated and various testing including mechanical and chemistry analysis were performed in order to evaluate the integrity of Yonggwang unit 2 reactor vessel during the operation until life time. The evaluation results are as follows; Fast neutron fluences for capsules U, V, X and W are 5.762E+18, 1.5391E+19, 3.5119E+19, and 4.2610E+19 n/cm{sup 2}, respectively. The bias factor, the ratio of measurement versus calculation, was 0.899 for the 1st through 4th testing and the calculational uncertainty, 12.3% satisfied the requirement of USNRC Reg.Guide DG-1053, 20%. The best estimated neutron fluence for reactor vessel inside surface was 1.357E+19 n/cm{sup 2} based on the end of 11th fuel cycle and it was predicted that the fluences of vessel inside surface at 32, 40, 48 and 56EFPY would reach 3.525E+19, 4.337E+19, 5.148E+19 and 5.960E+19 n/cm{sup 2} based on the current calculation. The result through this analysis for Yonggwang unit 2 showed that there would be no problem for the pressurized thermal shock(PTS) during the operation until design life. 48 refs., 35 figs., 42 tabs. (Author)

  8. Main mechanisms of material properties degradation under reactor pressure vessel operating conditions

    Energy Technology Data Exchange (ETDEWEB)

    Karzov, Georgy; Timofeev, Boris [Central Research Inst. of Structural Materials ' prometey' , St. Petersburg (Russian Federation)

    1999-07-01

    In the process of NPP equipment operation materials are subjected to a prolonged influence of loads, associated with the variation of inner pressure and temperature under various conditions. Each equipment element damage is associated with some material fracture mechanism. For NPP equipment the mechanisms of irreversible damage accumulation are related with: irradiation embrittlement, thermal and strain aging, fatigue damages from mechanical and thermal loading, stress corrosion and fatigue corrosion, creep and thermal relaxation stresses, erosion and weak, thermal shock. The basic tasks of specialists working in the sphere of the provision of reliability and service life of nuclear power equipment are not only the determination of the main mechanisms of damages and reasons of their appearance, but also the study of methods which would permit to control these properties completely. By giving some examples of Russian NPP equipment with VVER-440 and VVER-1000 reactors the paper presents most typical degradation mechanisms of equipment material properties, including weldments, in the process of operation and methods to recover by using various technological means. (author)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-05-15

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

  10. 1-Dimensional simulation of thermal annealing in a commercial nuclear power plant reactor pressure vessel wall section

    Energy Technology Data Exchange (ETDEWEB)

    Nakos, J.T.; Rosinski, S.T.; Acton, R.U.

    1994-11-01

    The objective of this work was to provide experimental heat transfer boundary condition and reactor pressure vessel (RPV) section thermal response data that can be used to benchmark computer codes that simulate thermal annealing of RPVS. This specific protect was designed to provide the Electric Power Research Institute (EPRI) with experimental data that could be used to support the development of a thermal annealing model. A secondary benefit is to provide additional experimental data (e.g., thermal response of concrete reactor cavity wall) that could be of use in an annealing demonstration project. The setup comprised a heater assembly, a 1.2 in {times} 1.2 m {times} 17.1 cm thick [4 ft {times} 4 ft {times} 6.75 in] section of an RPV (A533B ferritic steel with stainless steel cladding), a mockup of the {open_quotes}mirror{close_quotes} insulation between the RPV and the concrete reactor cavity wall, and a 25.4 cm [10 in] thick concrete wall, 2.1 in {times} 2.1 in [10 ft {times} 10 ft] square. Experiments were performed at temperature heat-up/cooldown rates of 7, 14, and 28{degrees}C/hr [12.5, 25, and 50{degrees}F/hr] as measured on the heated face. A peak temperature of 454{degrees}C [850{degrees}F] was maintained on the heated face until the concrete wall temperature reached equilibrium. Results are most representative of those RPV locations where the heat transfer would be 1-dimensional. Temperature was measured at multiple locations on the heated and unheated faces of the RPV section and the concrete wall. Incident heat flux was measured on the heated face, and absorbed heat flux estimates were generated from temperature measurements and an inverse heat conduction code. Through-wall temperature differences, concrete wall temperature response, heat flux absorbed into the RPV surface and incident on the surface are presented. All of these data are useful to modelers developing codes to simulate RPV annealing.

  11. 1-Dimensional simulation of thermal annealing in a commercial nuclear power plant reactor pressure vessel wall section

    Energy Technology Data Exchange (ETDEWEB)

    Nakos, J.T.; Rosinski, S.T.; Acton, R.U.

    1994-11-01

    The objective of this work was to provide experimental heat transfer boundary condition and reactor pressure vessel (RPV) section thermal response data that can be used to benchmark computer codes that simulate thermal annealing of RPVS. This specific protect was designed to provide the Electric Power Research Institute (EPRI) with experimental data that could be used to support the development of a thermal annealing model. A secondary benefit is to provide additional experimental data (e.g., thermal response of concrete reactor cavity wall) that could be of use in an annealing demonstration project. The setup comprised a heater assembly, a 1.2 in {times} 1.2 m {times} 17.1 cm thick [4 ft {times} 4 ft {times} 6.75 in] section of an RPV (A533B ferritic steel with stainless steel cladding), a mockup of the {open_quotes}mirror{close_quotes} insulation between the RPV and the concrete reactor cavity wall, and a 25.4 cm [10 in] thick concrete wall, 2.1 in {times} 2.1 in [10 ft {times} 10 ft] square. Experiments were performed at temperature heat-up/cooldown rates of 7, 14, and 28{degrees}C/hr [12.5, 25, and 50{degrees}F/hr] as measured on the heated face. A peak temperature of 454{degrees}C [850{degrees}F] was maintained on the heated face until the concrete wall temperature reached equilibrium. Results are most representative of those RPV locations where the heat transfer would be 1-dimensional. Temperature was measured at multiple locations on the heated and unheated faces of the RPV section and the concrete wall. Incident heat flux was measured on the heated face, and absorbed heat flux estimates were generated from temperature measurements and an inverse heat conduction code. Through-wall temperature differences, concrete wall temperature response, heat flux absorbed into the RPV surface and incident on the surface are presented. All of these data are useful to modelers developing codes to simulate RPV annealing.

  12. Fracture mechanics characterisation of the WWER-440 reactor pressure vessel beltline welding seam of Greifswald unit 8

    Energy Technology Data Exchange (ETDEWEB)

    Viehrig, Hans-Werner; Schuhknecht, Jan [Forschungszentrum Dresden-Rossendorf, Dresden (Germany)

    2009-07-01

    Russian type WWER reactors are operated in Russia and many other European countries like Finland, Czech Republic, Slovak Republic, Hungary, Bulgaria and Ukraine. Surveillance specimen programmes for the inspection of the aging of the reactor pressure vessel (RPV) materials were implemented for the second generation of WWER-440/V-213 reactors. The test results and the RPV integrity assessment has been evaluated according to national codes based on the Russian code PNAE G-7-002-86 ''Strength Calculation Norms for Nuclear Power Plant Equipment and Piping'' [1]. This is an indirect and correlative approach of determining the fracture toughness of the RPV steels in the initial and irradiated condition. The Master Curve (MC) approach as adopted in the test procedure ASTM E1921 [2] for assessing the fracture toughness of sampled irradiated materials has been gaining acceptance throughout the world [3]. The MC approach is more naturally suited to probabilistic analyses because it defines both a mean transition toughness value and a distribution around that value. It contains the assumptions of macroscopically homogenous material with uniform distribution of crack initiating defects along the crack front. In contrast to present indirect and correlative approach the specimen orientation and especially the crack extension direction in multilayer weld metal becomes more important for the direct measurement of the fracture toughness with Charpy size SE(B) specimens. The orientation of the Charpy- and SE(B) specimens is different for RPVs manufactured in Russia and by the SKODA company in the former Czechoslovakia [4,5]. Particularly with regard to weld metal it can be expected that the parameters of fracture toughness measured with Charpy-V or SE(B) specimens are strongly influenced by the specimen orientation. It raises the question whether the MC approach is also applicable when the structure varies along the crack front which is happen in TL oriented SE

  13. Characterisation of interfacial segregation to Cu-enriched precipitates in two thermally aged reactor pressure vessel steel welds

    Energy Technology Data Exchange (ETDEWEB)

    Styman, P.D., E-mail: paul.styman@materials.ox.ac.uk [Department of Materials, University of Oxford, Parks Road, Oxford OX1 3PH (United Kingdom); National Nuclear Laboratory, B168, Harwell Business Centre, Didcot, Oxon OX11 0QT (United Kingdom); Hyde, J.M. [Department of Materials, University of Oxford, Parks Road, Oxford OX1 3PH (United Kingdom); National Nuclear Laboratory, B168, Harwell Business Centre, Didcot, Oxon OX11 0QT (United Kingdom); Wilford, K.; Parfitt, D.; Riddle, N. [Rolls-Royce, PO Box 2000, Raynesway, Derby DE21 7XX (United Kingdom); Smith, G.D.W. [Department of Materials, University of Oxford, Parks Road, Oxford OX1 3PH (United Kingdom)

    2015-12-15

    To understand the contribution of long term thermal ageing to Reactor Pressure Vessel (RPV) embrittlement two high Cu steel welds with different Ni contents were thermally aged for times up to 100,000 h at 330 °C and 365 °C. Microstructural characterisation using Atom Probe Tomography was performed. Thermal ageing produced a high number density of nano-scale Cu-enriched precipitates. The precipitate–matrix interfaces were enriched in Ni, Mn and Si. The characterisation of these interfaces using a double cluster search approach is the subject of this work. The interface region around thermally-induced precipitates was found to be wider in steels with higher bulk Ni contents and where precipitates had larger core radii. The effect of ageing temperature on interface width was small when comparing precipitates of equal core radius. The narrower interface width in the lower Ni steels is reflected in the composition of the interface, which has a lower Ni content than in the higher Ni material. The reduction in interfacial energy due to the segregation of Ni, Mn and Si has been calculated and shows enhanced reductions in interfacial energy with increasing precipitate size, but no obvious effect of temperature. - Highlights: • Characterisation of interfacial segregation of Ni, Mn and Si to Cu-enriched clusters. • Analysis method gives information on interface composition and widths of large numbers of clusters. • Reduction in interface energy due to segregation of Ni, Mn and Si is calculated.

  14. Temperature distributions in a Tokamak vacuum vessel of fusion reactor after the loss-of-vacuum-events occurred

    Energy Technology Data Exchange (ETDEWEB)

    Takase, K.; Kunugi, T.; Shibata, M.; Seki, Y. [Japan Atomic Energy Research Inst., Naka, Ibaraki (Japan)

    1998-09-01

    If a loss-of-vacuum-event (LOVA) occurred in a fusion reactor, buoyancy-driven exchange flows would occur at breaches of a vacuum vessel (VV) due to the temperature difference between the inside and outside of the VV. The exchange flows may bring mixtures of activated materials and tritium in the VV to the outside through the breaches, and remove decay heat from the plasma-facing components of the VV. Therefore, the LOVA experiments were carried out under the condition that one or two breaches was opened and that the VV was heated to a maximum 200 C, using a small-scaled LOVA experimental apparatus. Air and helium gas were provided as working fluids. Fluid and wall temperature distributions in the VV were measured and the flow patterns in the VV were estimated by using these temperature distributions. It was found that: (1) the exchange mass in the VV depended on the breach positions; (2) the exchange flow at the single breach case became a counter-current flow when the breach was at the roof of the VV and a stratified flow when it was at the side wall; (3) and that at the double breach case, a one-way flow between two breaches was formed. (orig.) 6 refs.

  15. Microstructural behavior of VVER-440 reactor pressure vessel steels under irradiation to neutron fluences beyond the design operation period

    Science.gov (United States)

    Kuleshova, E. A.; Gurovich, B. A.; Shtrombakh, Ya. I.; Nikolaev, Yu. A.; Pechenkin, V. A.

    2005-06-01

    Electron-microscopy and fractographic studies of the surveillance specimens from base and weld metals of VVER-440/213 reactor pressure vessel (RPV) in the original state and after irradiations to different fast neutron fluences from ˜5 × 10 23 n m -2 ( E > 0.5 MeV) up to over design values have been carried out. The maximum specimens irradiation time was 84 480 h. It is shown that there is an evolution in radiation-induced structural behavior with radiation dose increase, which causes a change in relative contribution of the mechanisms responsible for radiation embrittlement of RPV materials. Particularly, radiation coalescence of copper-enriched precipitates and extensive density increase of dislocation loops was observed. Increase in dislocation loop density was shown to provide the dominant contribution to radiation hardening at the late irradiation stages (after reaching double the design end-of-life neutron fluence of ˜4 × 10 24 n m -2). The fracture mechanism of the base metal at those stages was observed to change from transcrystalline to intercrystalline.

  16. Neutron fluence at the reactor pressure vessel wall - a comparison of French and German procedures and strategies in PWRs

    Energy Technology Data Exchange (ETDEWEB)

    Tricot, N. [Institut de Radioprotection et de Surete Nucleaire, IRSN/DES/SECCA, 92 - Fontenay aux Roses (France); Jendrich, U. [Gesellschaft fuer Anlagen- und Reaktorsicherheit (GRS) mbH, Garching (Germany)

    2003-01-01

    While the neutrons within the core may take part in the chain reaction, those neutrons emitted from the core are basically lost for the energy production. This 'neutron leakage' represents a loss of fuel efficiency and causes neutron embrittlement of the reactor pressure vessel (RPV) wall. The latter raises safety concerns, needs to be monitored closely and may necessitate mitigating measures. There are different strategies to deal with these two undesirable effects: The neutron emission may be reduced to some extent all around the core or just at the 'hot spots' of RPV embrittlement by tailored core loading patterns. A higher absorption rate of neutrons may also be achieved by a larger water gap between the core and the RPV. In this paper the inter-relations between the distribution of neutron flux, core geometry, core loading strategy, RPV embrittlement and its surveillance are discussed at first. Then the different strategies followed by the German and French operators are described. Finally the conclusions will highlight the communalities and differences between these strategies as different approaches to the same problem of safety as well as economy. (authors)

  17. Microstructural changes of a thermally aged stainless steel submerged arc weld overlay cladding of nuclear reactor pressure vessels

    Energy Technology Data Exchange (ETDEWEB)

    Takeuchi, T., E-mail: takeuchi.tomoaki@jaea.go.jp [Japan Atomic Energy Agency, Tokai, Ibaraki 319-1195 (Japan); Kameda, J. [National Institute for Materials Science, Sengen, Tsukuba 305-0047 (Japan); Nagai, Y.; Toyama, T.; Matsukawa, Y. [Oarai Center, Institute for Materials Research, Tohoku University, Oarai, Ibaraki 311-1313 (Japan); Nishiyama, Y.; Onizawa, K. [Japan Atomic Energy Agency, Tokai, Ibaraki 319-1195 (Japan)

    2012-06-15

    The effect of thermal aging on microstructural changes in stainless steel submerged arc weld-overlay cladding of reactor pressure vessels was investigated using atom probe tomography (APT). In as-received materials subjected to post-welding heat treatments (PWHTs), with a subsequent furnace cooling, a slight fluctuation of the Cr concentration was observed due to spinodal decomposition in the {delta}-ferrite phase but not in the austenitic phase. Thermal aging at 400 Degree-Sign C for 10,000 h caused not only an increase in the amplitude of spinodal decomposition but also the precipitation of G phases with composition ratios of Ni:Si:Mn = 16:7:6 in the {delta}-ferrite phase. The degree of the spinodal decomposition in the submerged arc weld sample was similar to that in the electroslag weld one reported previously. We also observed a carbide on the {gamma}-austenite and {delta}-ferrite interface. There were no Cr depleted zones around the carbide.

  18. Development and evaluation of thermoelectric power measurements as a non destructive technique to evaluate ageing of reactor pressure vessel steels

    Energy Technology Data Exchange (ETDEWEB)

    Acosta, B.; Debarberis, L. [European Commission, JRC Institute for Advanced Materials, Petten (Netherlands); Perlado, J.M. [Universidad Politecnica de Madrid, Instituto de Fusion Nuclear, DENIM, E.T.S.I.I., Madris (Spain)

    2001-07-01

    The STEAM (Seebeck and Thomson effects on aged materials) technique developed at the JRC-IAM (joint research centre - institute for advanced materials), is a new non-destructive method able to detect in a simple way degradation of materials, in particular to be applied on those steels that form the reactor pressure vessel of nuclear plants. The STEAM method is based on the measurement of the thermoelectric voltage generated by the Seebeck and Thomson effects taking place in the material under test. In order to evaluate the performance of the STEAM technique on irradiated material a set of 32 model alloys was selected. Measurements with the STEAM technique have been performed on the model alloys in both conditions, fresh and irradiated, with the aim of correlating the irradiation induced embrittlement and the change on the Seebeck coefficient due to irradiation. The results show that there is a relationship between transition temperature shifts and Seebeck coefficient value change between irradiated and fresh materials. In order to understand the response of the Seebeck coefficient to neutron irradiation damage a model based on multivariable correlation analysis is proposed. (A.C.)

  19. An object kinetic Monte Carlo model for the microstructure evolution of neutron-irradiated reactor pressure vessel steels

    Energy Technology Data Exchange (ETDEWEB)

    Messina, Luca; Olsson, Paer [KTH Royal Institute of Technology, Stockholm (Sweden); Chiapetto, Monica [SCK - CEN, Nuclear Materials Science Institute, Mol (Belgium); Unite Materiaux et Transformations (UMET), UMR 8207, Universite de Lille 1, ENSCL, Villeneuve d' Ascq (France); Becquart, Charlotte S. [Unite Materiaux et Transformations (UMET), UMR 8207, Universite de Lille 1, ENSCL, Villeneuve d' Ascq (France); Malerba, Lorenzo [SCK - CEN, Nuclear Materials Science Institute, Mol (Belgium)

    2016-11-15

    This work presents a full object kinetic Monte Carlo framework for the simulation of the microstructure evolution of reactor pressure vessel (RPV) steels. The model pursues a ''gray-alloy'' approach, where the effect of solute atoms is seen exclusively as a reduction of the mobility of defect clusters. The same set of parameters yields a satisfactory evolution for two different types of alloys, in very different irradiation conditions: an Fe-C-MnNi model alloy (high flux) and a high-Mn, high-Ni RPV steel (low flux). A satisfactory match with the experimental characterizations is obtained only if assuming a substantial immobilization of vacancy clusters due to solute atoms, which is here verified by means of independent atomistic kinetic Monte Carlo simulations. The microstructure evolution of the two alloys is strongly affected by the dose rate; a predominance of single defects and small defect clusters is observed at low dose rates, whereas larger defect clusters appear at high dose rates. In both cases, the predicted density of interstitial loops matches the experimental solute-cluster density, suggesting that the MnNi-rich nanofeatures might form as a consequence of solute enrichment on immobilized small interstitial loops, which are invisible to the electron microscope. (copyright 2016 WILEY-VCH Verlag GmbH and Co. KGaA, Weinheim)

  20. Advanced Hydride Laboratory

    Energy Technology Data Exchange (ETDEWEB)

    Motyka, T.

    1989-01-01

    Metal hydrides have been used at the Savannah River Tritium Facilities since 1984. However, the most extensive application of metal hydride technology at the Savannah River Site is being planned for the Replacement Tritium Facility, a $140 million facility schedules for completion in 1990 and startup in 1991. In the new facility, metal hydride technology will be used to store, separate, isotopically purify, pump, and compress hydrogen isotopes. In support of the Replacement Tritium Facility, a $3.2 million, cold,'' process demonstration facility, the Advanced Hydride Laboratory began operation in November of 1987. The purpose of the Advanced Hydride Laboratory is to demonstrate the Replacement Tritium Facility's metal hydride technology by integrating the various unit operations into an overall process. This paper will describe the Advanced Hydride Laboratory, its role and its impact on the application of metal hydride technology to tritium handling.

  1. Advanced Hydride Laboratory

    Energy Technology Data Exchange (ETDEWEB)

    Motyka, T.

    1989-12-31

    Metal hydrides have been used at the Savannah River Tritium Facilities since 1984. However, the most extensive application of metal hydride technology at the Savannah River Site is being planned for the Replacement Tritium Facility, a $140 million facility schedules for completion in 1990 and startup in 1991. In the new facility, metal hydride technology will be used to store, separate, isotopically purify, pump, and compress hydrogen isotopes. In support of the Replacement Tritium Facility, a $3.2 million, ``cold,`` process demonstration facility, the Advanced Hydride Laboratory began operation in November of 1987. The purpose of the Advanced Hydride Laboratory is to demonstrate the Replacement Tritium Facility`s metal hydride technology by integrating the various unit operations into an overall process. This paper will describe the Advanced Hydride Laboratory, its role and its impact on the application of metal hydride technology to tritium handling.

  2. 核电压力容器焊接过程中的生产管理%Production Control to Welding of Nuclear Reactor Pressure Vessels

    Institute of Scientific and Technical Information of China (English)

    宋桂艳

    2014-01-01

    The article describes the welding method for nuclear reactor pressure vessels and the key points of plan control and welding production control and process control in welding operation.%概述核电压力容器焊接的特点,指出焊接生产过程中计划管理、焊接生产过程管理和过程控制的重点内容。

  3. Installation of Reactor Vessel in AP1000 Nuclear Unit%AP1000核电机组反应堆压力容器的安装

    Institute of Scientific and Technical Information of China (English)

    许跃武; 高宝宁

    2012-01-01

    Hoisting orientation, rotation and location adjustment of reactor vessel in Sanmen Power Unit 1 were illustrated. By calculating the hoisting height, weight and the wind load, analyzed the safety of "open top" way on lifting reactor vessel. The methods to measure and adjust the location, levelness and elevation were also introduced, which after reactor vessel in place. Furthermore, the installation of supports and keeper plates were discussed. It supply reference value for the same nuclear plant in the hoisting and installation of main components.%阐述了某核电1号机组反应堆压力容器吊装定位、筒体翻转和就位调整的过程.通过对吊装高度、重量及风载荷进行计算,分析“开顶法”吊装反应堆压力容器的安全性.介绍了反应堆压力容器就位后方位、水平度、标高的测量与调整方法,以及护板、支撑的安装.对同类核电站的主设备吊装具有一定参考价值.

  4. Neutron radiation embrittlement studies in support of continued operation, and validation by sampling of Magnox reactor steel pressure vessels and components

    Energy Technology Data Exchange (ETDEWEB)

    Jones, R.B.; Bolton, C.J. [Magnox Electric plc, Berkeley Centre, Glos (United Kingdom)

    1997-02-01

    Magnox steel reactor pressure vessels differ significantly from US LWR vessels in terms of the type of steel used, as well as their operating environment (dose level, exposure temperature range, and neutron spectra). The large diameter ferritic steel vessels are constructed from C-Mn steel plates and forgings joined together with manual metal and submerged-arc welds which are stress-relieved. All Magnox vessels are now at least thirty years old and their continued operation is being vigorously pursued. Vessel surveillance and other programmes are summarized which support this objective. The current understanding of the roles of matrix irradiation damage, irradiation-enhanced copper impurity precipitation and intergranular embrittlement effects is described in so far as these influence the form of the embrittlement and hardening trend curves for each material. An update is given on the influence of high temperature exposure, and on the role of differing neutron spectra. Finally, the validation offered by the results of an initial vessel sampling exercise is summarized together with the objectives of a more extensive future sampling programme.

  5. Development of Mini-Compact Tension Test Method for Determining Fracture Toughness Master Curves for Reactor Pressure Vessel Steels

    Energy Technology Data Exchange (ETDEWEB)

    Sokolov, Mikhail A. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2017-05-01

    Small specimens are playing the key role in evaluating properties of irradiated materials. The use of small specimens provides several advantages. Typically, only a small volume of material can be irradiated in a reactor at desirable conditions in terms of temperature, neutron flux, and neutron dose. A small volume of irradiated material may also allow for easier handling of specimens. Smaller specimens reduce the amount of radioactive material, minimizing personnel exposures and waste disposal. However, use of small specimens imposes a variety of challenges as well. These challenges are associated with proper accounting for size effects and transferability of small specimen data to the real structures of interest. Any fracture toughness specimen that can be made out of the broken halves of standard Charpy specimens may have exceptional utility for evaluation of reactor pressure vessels (RPVs) since it would allow one to determine and monitor directly actual fracture toughness instead of requiring indirect predictions using correlations established with impact data. The Charpy V-notch specimen is the most commonly used specimen geometry in surveillance programs. Validation of the mini compact tension specimen (mini-CT) geometry has been performed on previously well characterized Midland beltline Linde 80 (WF-70) weld in the unirradiated condition. It was shown that the fracture toughness transition temperature, To, measured by these Mini-CT specimens is almost the same as To value that was derived from various larger fracture toughness specimens. Moreover, an International collaborative program has been established to extend the assessment and validation efforts to irradiated Linde 80 weld metal. The program is underway and involves the Oak Ridge National Laboratory (ORNL), Central Research Institute for Electrical Power Industry (CRIEPI), and Electric Power Research Institute (EPRI). The irradiated Mini-CT specimens from broken halves of previously tested Charpy

  6. Nanostructure evolution under irradiation of Fe(C)MnNi model alloys for reactor pressure vessel steels

    Science.gov (United States)

    Chiapetto, M.; Becquart, C. S.; Domain, C.; Malerba, L.

    2015-06-01

    Radiation-induced embrittlement of bainitic steels is one of the most important lifetime limiting factors of existing nuclear light water reactor pressure vessels. The primary mechanism of embrittlement is the obstruction of dislocation motion produced by nanometric defect structures that develop in the bulk of the material due to irradiation. The development of models that describe, based on physical mechanisms, the nanostructural changes in these types of materials due to neutron irradiation are expected to help to better understand which features are mainly responsible for embrittlement. The chemical elements that are thought to influence most the response under irradiation of low-Cu RPV steels, especially at high fluence, are Ni and Mn, hence there is an interest in modelling the nanostructure evolution in irradiated FeMnNi alloys. As a first step in this direction, we developed sets of parameters for object kinetic Monte Carlo (OKMC) simulations that allow this to be done, under simplifying assumptions, using a "grey alloy" approach that extends the already existing OKMC model for neutron irradiated Fe-C binary alloys [1]. Our model proved to be able to describe the trend in the buildup of irradiation defect populations at the operational temperature of LWR (∼300 °C), in terms of both density and size distribution of the defect cluster populations, in FeMnNi model alloys as compared to Fe-C. In particular, the reduction of the mobility of point-defect clusters as a consequence of the presence of solutes proves to be key to explain the experimentally observed disappearance of detectable point-defect clusters with increasing solute content.

  7. Fracture mechanics characterisation of the WWER-440 reactor pressure vessel beltline welding seam of Greifswald unit 8

    Energy Technology Data Exchange (ETDEWEB)

    Viehrig, Hans-Werner; Schuhknecht, Jan [Forschungszentrum Dresden-Rossendorf (Germany)

    2008-07-01

    WWER-440 second generation (V-213) reactor pressure vessels (RPV) were produced by IZHORA in Russia and by SKODA in the former Czechoslovakia. The surveillance Charpy-V and fracture mechanics SE(B) specimens of both producers have different orientations. The main difference is the crack extension direction which is through the RPV thickness and circumferential for ISHORA and SKODA RPV, respectively. In particular for the investigation of weld metal from multilayer submerged welding seams the crack extension direction is of importance. Depending on the crack extension direction in the specimen there are different welding beads or a uniform structure along the crack front. The specimen orientation becomes more important when the fracture toughness of the weld metal is directly determined on surveillance specimens according to the Master Curve (MC) approach as standardised in the ASTM Standard Test Method E1921. This approach was applied on weld metal of the RPV beltline welding seam of Greifswald Unit 8 RPV. Charpy size SE(B) specimens from 13 locations equally spaced over the thickness of the welding seam were tested. The specimens are in TL and TS orientation. The fracture toughness values measured on the SE(B) specimens with both orientations follow the course of the MC. Nearly all values lie within the fracture toughness curves for 5% and 95% fracture probability. There is a strong variation of the reference temperature T{sub 0} though the thickness of the welding seam, which can be explained with structural differences. The scatter is more pronounced for the TS SE(B) specimens. It can be shown that specimens with TS and TL orientation in the welding seam have a differentiating and integrating behaviour, respectively. The statistical assumptions behind the MC approach are valid for both specimen orientations even if the structure is not uniform along the crack front. By comparison crack extension, JR, curves measured on SE(B) specimens with TL and TS orientation

  8. Assessment of segregation kinetics in water-moderated reactors pressure vessel steels under long-term operation

    Science.gov (United States)

    Kuleshova, E. A.; Gurovich, B. A.; Lavrukhina, Z. V.; Saltykov, M. A.; Fedotova, S. V.; Khodan, A. N.

    2016-08-01

    In reactor pressure vessel (RPV) bcc-lattice steels temper embrittlement is developed under the influence of both operating temperature of ∼300 °C and neutron irradiation. Segregation processes in the grain boundaries (GB) begin to play a special role in the assessment of the safe operation of the RPV in case of its lifetime extension up to 60 years or more. The most reliable information on the RPV material condition can be obtained by investigating the surveillance specimens (SS) that are exposed to operational factors simultaneously with the RPV itself. In this paper the GB composition in the specimens with different thermal exposure time at the RPV operating temperature as well as irradiated by fast neutrons (E ≥ 0.5 MeV) to different fluences (20-71)·1022 m-2 was studied by means of Auger electron spectroscopy (AES) including both impurity and main alloying elements content. The data obtained allowed to trace the trend of the operating temperature and radiation-stimulated diffusion influence on the overall segregants level in GB. The revealed differences in the concentration levels of GB segregants in different steels, are due to the different chemical composition of the steels and also due to different grain boundary segregation levels in initial (unexposed) state. The data were used to estimate the RPV steels working capacity for 60 years. The estimation was carried out using both the well-known Langmuir-McLean model and the one specially developed for RPV steels, which takes into account the structure and phase composition of VVER-1000 RPV steels, as well as the long-term influence of operational factors.

  9. The effect of non-metallic inclusions on the fracture toughness master curve in high copper reactor pressure vessel welds

    Science.gov (United States)

    Oh, Yong-Jun; Lee, Bong-Sang; Hong, Jun-Hwa

    2002-03-01

    The fracture toughness of two high copper reactor pressure vessel welds having low upper shelf energy was evaluated in accordance with the master curve method of ASTM E1921. The resultant data were correlated to the metallurgical factors involved in the brittle fracture initiation to provide a metallurgical-based understanding of the master curve. The tests were performed using pre-cracked Charpy V-notched specimens and the master curve was made with an average of T0 values determined at different temperatures. In all specimens, the cleavage fracture initiated at non-metallic inclusion ranging from 0.7 to 3.5 μm in diameter showing a scatter with the specimens and testing temperatures. Temperature dependency of the triggering particle size was not found. The fracture toughness ( KJC) was inversely proportional to the square root of the triggering inclusion diameter ( di) at respective temperatures. From this relationship, we determined median KJC values which correspond to the average value of triggering inclusion diameter of all tested specimens and defined them as a modified median KJC ( K'JC(med) ). The obtained K'JC(med) values showed quite smaller deviation from the master curve at different temperatures than the experimental median KJC values. This suggests that the master curve is on the premise of a constant dimension of key microstructural factor in a material regardless of the testing temperature. But the inclusion size at trigger point played an important role in the absolute position of the master curve with temperature and the consequent T0 value.

  10. Microstructural parameters governing cleavage fracture behaviors in the ductile-brittle transition region in reactor pressure vessel steels

    Energy Technology Data Exchange (ETDEWEB)

    Yang, Won-Jon; Lee, Bong-Sang; Oh, Yong-Jun; Huh, Moo-Young; Hong, Jun-Hwa

    2004-08-15

    The fracture behaviors in the ductile-brittle transition region of reactor pressure vessel (RPV) steels with similar chemical compositions but different manufacturing processes were examined in view of cleavage fracture stress at crack-tip. The steels typically had a variation in grain size and carbide size distribution through the different manufacturing processes. Fracture toughness was evaluated by using a statistical method in accordance to the ASTM standard E1921. From the fractography of the tested specimens, it was found that fracture toughness of the steels increased with increasing distance from the crack-tip to the cleavage initiating location, namely cleavage initiation distance (CID, X{sub f}) and its statistical mean value (K{sub JC(med)}) was proportional to the cleavage fracture stress ({sigma}{sub f}) determined from finite-element (FE) calculation at cleavage initiating location. On the other hand, {sigma}{sub f} could also be calculated by applying the size of microstructural parameters, such as carbide, grain and bainite packet, into the Griffith's theory for brittle fracture. Among the parameters, the {sigma}{sub f} obtained from the mean diameter of the carbides above 1% of the total population was in good agreement with the {sigma}{sub f} value from the FE calculation for the five different steels. The results suggest that the fracture toughness of bainitic RPV steels in the transition region is mostly influenced by only some 1% of total carbides and the critical step for cleavage fracture of the RPV steels should be the propagation of this carbide size crack to the adjacent ferrite matrix.

  11. The effect of non-metallic inclusions on the fracture toughness master curve in high copper reactor pressure vessel welds

    Energy Technology Data Exchange (ETDEWEB)

    Oh, Yong-Jun E-mail: yjoh@kaeri.re.kr; Lee, Bong-Sang; Hong, Jun-Hwa

    2002-03-01

    The fracture toughness of two high copper reactor pressure vessel welds having low upper shelf energy was evaluated in accordance with the master curve method of ASTM E1921. The resultant data were correlated to the metallurgical factors involved in the brittle fracture initiation to provide a metallurgical-based understanding of the master curve. The tests were performed using pre-cracked Charpy V-notched specimens and the master curve was made with an average of T{sub 0} values determined at different temperatures. In all specimens, the cleavage fracture initiated at non-metallic inclusion ranging from 0.7 to 3.5 {mu}m in diameter showing a scatter with the specimens and testing temperatures. Temperature dependency of the triggering particle size was not found. The fracture toughness (K{sub J{sub C}}) was inversely proportional to the square root of the triggering inclusion diameter (d{sub i}) at respective temperatures. From this relationship, we determined median K{sub J{sub C}} values which correspond to the average value of triggering inclusion diameter of all tested specimens and defined them as a modified median K{sub J{sub C}} (K{sup '}{sub J{sub C}}{sub (med)}). The obtained K{sup '}{sub J{sub C}}{sub (med)} values showed quite smaller deviation from the master curve at different temperatures than the experimental median K{sub J{sub C}} values. This suggests that the master curve is on the premise of a constant dimension of key microstructural factor in a material regardless of the testing temperature. But the inclusion size at trigger point played an important role in the absolute position of the master curve with temperature and the consequent T{sub 0} value.

  12. Attrition reactor system

    Science.gov (United States)

    Scott, Charles D.; Davison, Brian H.

    1993-01-01

    A reactor vessel for reacting a solid particulate with a liquid reactant has a centrifugal pump in circulatory flow communication with the reactor vessel for providing particulate attrition, resulting in additional fresh surface where the reaction can occur.

  13. High Fluency Low Flux Embrittlement Models of LWR Reactor Pressure Vessel Embrittlement and a Supporting Database from the UCSB ATR-2 Irradiation Experiment

    Energy Technology Data Exchange (ETDEWEB)

    Odette, G. Robert [Univ. of California, Santa Barbara, CA (United States)

    2017-01-24

    Reactor pressure vessel embrittlement may limit the lifetime of light water reactors (LWR). Embrittlement is primarily caused by formation of nano-scale precipitates, which cause hardening and a subsequent increase in the ductile-to-brittle transition temperature of the steel. While the effect of Cu has historically been the largest research focus of RPV embrittlement, there is increasing evidence that Mn, Ni and Si are likely to have a large effect at higher fluence, where Mn-Ni-Si precipitates can form, even in the absence of Cu. Therefore, extending RPV lifetimes will require a thorough understanding of both precipitation and embrittlement at higher fluences than have ever been observed in a power reactor. To address this issue, test reactors that irradiate materials at higher neutron fluxes than power reactors are used. These experiments at high neutron flux can reach extended life neutron fluences in only months or several years. The drawback of these test irradiations is that they add additional complexity to interpreting the data, as the irradiation flux also plays a role into both precipitate formation and irradiation hardening and embrittlement. This report focuses on developing a database of both microstructure and mechanical property data to better understand the effect of flux. In addition, a previously developed model that enables the comparison of data taken over a range of neutron flux is discussed.

  14. Argonne Liquid-Metal Advanced Burner Reactor : components and in-vessel system thermal-hydraulic research and testing experience - pathway forward.

    Energy Technology Data Exchange (ETDEWEB)

    Kasza, K.; Grandy, C.; Chang, Y.; Khalil, H.; Nuclear Engineering Division

    2007-06-30

    This white paper provides an overview and status report of the thermal-hydraulic nuclear research and development, both experimental and computational, conducted predominantly at Argonne National Laboratory. Argonne from the early 1970s through the early 1990s was the Department of Energy's (DOE's) lead lab for thermal-hydraulic development of Liquid Metal Reactors (LMRs). During the 1970s and into the mid-1980s, Argonne conducted thermal-hydraulic studies and experiments on individual reactor components supporting the Experimental Breeder Reactor-II (EBR-II), Fast Flux Test Facility (FFTF), and the Clinch River Breeder Reactor (CRBR). From the mid-1980s and into the early 1990s, Argonne conducted studies on phenomena related to forced- and natural-convection thermal buoyancy in complete in-vessel models of the General Electric (GE) Prototype Reactor Inherently Safe Module (PRISM) and Rockwell International (RI) Sodium Advanced Fast Reactor (SAFR). These two reactor initiatives involved Argonne working closely with U.S. industry and DOE. This paper describes the very important impact of thermal hydraulics dominated by thermal buoyancy forces on reactor global operation and on the behavior/performance of individual components during postulated off-normal accident events with low flow. Utilizing Argonne's LMR expertise and design knowledge is vital to the further development of safe, reliable, and high-performance LMRs. Argonne believes there remains an important need for continued research and development on thermal-hydraulic design in support of DOE's and the international community's renewed thrust for developing and demonstrating the Global Nuclear Energy Partnership (GNEP) reactor(s) and the associated Argonne Liquid Metal-Advanced Burner Reactor (LM-ABR). This white paper highlights that further understanding is needed regarding reactor design under coolant low-flow events. These safety-related events are associated with the transition

  15. Nuclear reactor neutron shielding

    Energy Technology Data Exchange (ETDEWEB)

    Speaker, Daniel P; Neeley, Gary W; Inman, James B

    2017-09-12

    A nuclear reactor includes a reactor pressure vessel and a nuclear reactor core comprising fissile material disposed in a lower portion of the reactor pressure vessel. The lower portion of the reactor pressure vessel is disposed in a reactor cavity. An annular neutron stop is located at an elevation above the uppermost elevation of the nuclear reactor core. The annular neutron stop comprises neutron absorbing material filling an annular gap between the reactor pressure vessel and the wall of the reactor cavity. The annular neutron stop may comprise an outer neutron stop ring attached to the wall of the reactor cavity, and an inner neutron stop ring attached to the reactor pressure vessel. An excore instrument guide tube penetrates through the annular neutron stop, and a neutron plug comprising neutron absorbing material is disposed in the tube at the penetration through the neutron stop.

  16. Effects of Repair Weld of Reactor Pressure Vessel Upper Head Control Rod Drive Mechanism Penetration Nozzle on J-Groove Weldment Using Finite Element Analysis

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Ju Hee; Yoo, Sam Hyeon [Korea Military Academy, Seoul (Korea, Republic of); Kim, Yun Jae [Korea University, Seoul (Korea, Republic of)

    2014-06-15

    In pressurized water reactors, the upper head of the reactor pressure vessel (RPV) contains numerous control rod drive mechanism (CRDM) nozzles. These nozzles are fabricated by welding after being inserted into the RPV head with a room temperature shrink fit. The tensile residual stresses caused by this welding are a major factor in primary water stress corrosion cracking (PWSCC). Over the last 15 years, the incidences of cracking in alloy 600 CRDM nozzles have increased significantly. These cracks are caused by PWSCC and have been shown to be driven by the welding residual stresses and operational stresses in the weld region. Various measures are being sought to overcome these problems. The defects resulting from the welding process are often the cause of PWSCC acceleration. Therefore, any weld defects found in the RPV manufacturing process are immediately repaired by repair welding. Detailed finite element simulations for the Korea Nuclear Reactor Pressure Vessel were conducted in order to predict the magnitudes of the repair weld residual stresses in the tube materials.

  17. High H- ionic conductivity in barium hydride

    Science.gov (United States)

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

    2015-01-01

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

  18. Synthesis of ruthenium hydride

    Science.gov (United States)

    Kuzovnikov, M. A.; Tkacz, M.

    2016-02-01

    Ruthenium hydride was synthesized at a hydrogen pressure of about 14 GPa in a diamond-anvil cell. Energy-dispersive x-ray diffraction was used to monitor the ruthenium crystal structure as a function of hydrogen pressure up to 30 GPa. The hydride formation was accompanied by phase transition from the original hcp structure of the pristine metal to the fcc structure. Our results confirmed the theoretical prediction of ruthenium hydride formation under hydrogen pressure. The standard Gibbs free energy of the ruthenium hydride formation reaction was calculated assuming the pressure of decomposition as the equilibrium pressure.

  19. Thermal–mechanical stress analysis of pressurized water reactor pressure vessel with/without a preexisting crack under grid load following conditions

    Energy Technology Data Exchange (ETDEWEB)

    Mohanty, Subhasish, E-mail: smohanty@anl.gov; Soppet, William K.; Majumdar, Saurin; Natesan, Krishnamurti

    2016-12-15

    Highlights: • Use of intermittent renewable-energy source in power grid is becoming a trend. • Gird load-following can leads to variable power demand from Nuclear power plant. • Reactor components can be stressed differently under gird load-following mode. • Estimation of stress–strain state under grid load-following condition is essential. - Abstract: In this paper, we present thermal–mechanical stress analysis of a pressurized water reactor pressure vessel and its hot-leg and cold-leg nozzles. Results are presented from thermal and thermal–mechanical stress analysis under reactor heat-up, cool-down, and grid load-following conditions. Analysis results are given with and without the presence of preexisting crack in the reactor nozzle (axial crack in hot leg nozzle). From the model results it is found that the stress–strain states are significantly higher in case of presence of crack than without crack. The stress–strain state under grid load following condition are more realistic compared to the stress–strain state estimated assuming simplified transients.

  20. Mechanical properties and microstructural investigations of TIG welded 40 mm and 60 mm thick SS 316L samples for fusion reactor vacuum vessel applications

    Energy Technology Data Exchange (ETDEWEB)

    Buddu, Ramesh Kumar, E-mail: brkumar75@gmail.com; Chauhan, N.; Raole, P.M.

    2014-12-15

    Highlights: • Austenitic stainless steels (316L) of 40 mm and 60 mm thickness plates were joined by Tungsten Inert Gas welding (TIG) process which are probable materials for advanced fusion reactor vacuum vessel requirements. • Mechanical properties and detailed microstructure studies have been carried out for welded samples. • Fractography analysis of impact test specimens indicated ductile fracture mode in BM, HAZ and WZ samples. • Presence of delta ferrite phase was observed in the welded zone and ferrite number data was measured for the base and weld metal and was found high in welds. - Abstract: The development of advanced fusion reactors like DEMO will have various challenges in materials and fabrication. The vacuum vessel is important part of the fusion reactor. The double walled design for vacuum vessel with thicker stainless steel material (40–60 mm) has been proposed in the advanced fusion reactors like ITER. Different welding techniques will have to be used for such vacuum vessel development. The required mechanical, structural and other properties of stainless steels have to be maintained in these joining processes of components of various shapes and sizes in the form of plates, ribs, shells, etc. The present paper reports characterization of welding joints of SS316L plates with higher thicknesses like 40 mm and 60 mm, prepared using multi-pass Tungsten Inert Gas (TIG) welding process. The weld quality has been evaluated with non-destructive tests by X-ray radiography and ultrasonic methods. The mechanical properties like tensile, bend tests, Vickers hardness and impact fracture tests have been carried out for the weld samples. Tensile property test results indicate sound weld joints with efficiencies over 100%. Hardening was observed in the weld zone in non-uniform manner. Macro and microstructure studies have been carried out for Base Metal (BM), Heat Affected Zone (HAZ) and Weld Zone (WZ). Scanning Electron Microscopy (SEM) analysis carried

  1. New Dosimetric Interpretation of the DV50 Vessel-Steel Experiment Irradiated in the OSIRIS MTR Reactor Using the Monte-Carlo Code TRIPOLI-4®

    Directory of Open Access Journals (Sweden)

    Malouch Fadhel

    2016-01-01

    Full Text Available An irradiation program DV50 was carried out from 2002 to 2006 in the OSIRIS material testing reactor (CEA-Saclay center to assess the pressure vessel steel toughness curve for a fast neutron fluence (E > 1 MeV equivalent to a French 900-MWe PWR lifetime of 50 years. This program allowed the irradiation of 120 specimens out of vessel steel, subdivided in two successive irradiations DV50 n∘1 and DV50 n∘2. To measure the fast neutron fluence (E > 1 MeV received by specimens after each irradiation, sample holders were equipped with activation foils that were withdrawn at the end of irradiation for activity counting and processing. The fast effective cross-sections used in the dosimeter processing were determined with a specific calculation scheme based on the Monte-Carlo code TRIPOLI-3 (and the nuclear data ENDF/B-VI and IRDF-90. In order to put vessel-steel experiments at the same standard, a new dosimetric interpretation of the DV50 experiment has been performed by using the Monte-Carlo code TRIPOLI-4 and more recent nuclear data (JEFF3.1.1 and IRDF-2002. This paper presents a comparison of previous and recent calculations performed for the DV50 vessel-steel experiment to assess the impact on the dosimetric interpretation.

  2. Analysis of the micro-structural damages by neutronic irradiation of the steel of reactor vessels of the nuclear power plant of Laguna Verde. Characterization of the design steel; Analisis de los danos micro-estructurales por irradiacion neutronica del acero de la vasija de los reactores de la Central Nuclear de Laguna Verde. Caracterizacion del acero de diseno

    Energy Technology Data Exchange (ETDEWEB)

    Moranchel y Rodriguez, M.; Garcia B, A. [IPN, Escuela Superior de Fisica y Matematicas, Departamento de Fisica, Av. Luis Enrique Erro s/n, Unidad Profesional Adolfo Lopez Mateos, Col. Lindavista, 07738 Mexico D. F. (Mexico); Longoria G, L. C., E-mail: mmoranchel@ipn.m [ININ, Direccion de Investigacion Cientifica, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico)

    2010-09-15

    The vessel of a nuclear reactor is one of the safety barriers more important in the design, construction and operation of the reactor. If the vessel results affected to the grade of to have fracture and/or cracks it is very probable the conclusion of their useful life in order to guarantee the nuclear safety and the radiological protection of the exposure occupational personnel, of the public and the environment avoiding the exposition to radioactive sources. The materials of the vessel of a nuclear reactor are exposed continually to the neutronic irradiation that generates the same nuclear reactor. The neutrons that impact to the vessel have the sufficient energy to penetrate certain depth in function of the energy of the incident neutron until reaching the repose or to be absorbed by some nucleus. In the course of their penetration, the neutrons interact with the nuclei, atoms, molecules and with the same crystalline nets of the vessel material producing vacuums, interstitial, precipitate and segregations among other defects that can modify the mechanical properties of the steel. The steel A533-B is the material with which is manufactured the vessel of the nuclear reactors of nuclear power plant of Laguna Verde, is an alloy that, among other components, it contains atoms of Ni that if they are segregated by the neutrons impact this would favor to the cracking of the same vessel. This work is part of an investigation to analyze the micro-structural damages of the reactor vessels of the nuclear power plant of Laguna Verde due to the neutronic irradiation which is exposed in a continuous way. We will show the characterization of the design steel of the vessel, what offers a comprehension about their chemical composition, the superficial topography and the crystalline nets of the steel A533-B. It will also allow analyze the existence of precipitates, segregates, the type of crystalline net and the distances inter-plains of the design steel of the vessel. (Author)

  3. Neutron flux calculation and fluence in the encircling of the core and vessel of a reactor BWR; Calculo de flujo neutronico y fluencia en la envolvente del nucleo y la vasija de un reactor BWR

    Energy Technology Data Exchange (ETDEWEB)

    Martinez C, E.

    2011-07-01

    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-{theta} 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-{theta}, 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, {theta} 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{sup 2}s, at a height H 4 (239.07 cm) and angle 32.236{sup o} in the core shroud and 4.00 E + 09 n/cm{sup 2}s at a height H 4 and angle 35.27{sup o} in the inner wall of the reactor vessel, positions that are consistent to within {+-}10% over the ones reported in the literature. (Author)

  4. A scaling law for the local CHF on the external bottom side of a fully submerged reactor vessel

    Energy Technology Data Exchange (ETDEWEB)

    Cheung, F.B.; Haddad, K.H.; Liu, Y.C. [Pennsylvania State Univ., University Park, PA (United States)

    1997-02-01

    A scaling law for estimating the local critical heat flux on the outer surface of a heated hemispherical vessel that is fully submerged in water has been developed from the results of an advanced hydrodynamic CHF model for pool boiling on a downward facing curved heating surface. The scaling law accounts for the effects of the size of the vessel, the level of liquid subcooling, the intrinsic properties of the fluid, and the spatial variation of the local critical heat flux along the heating surface. It is found that for vessels with diameters considerably larger than the characteristic size of the vapor masses, the size effect on the local critical heat flux is limited almost entirely to the effect of subcooling associated with the local liquid head. When the subcooling effect is accounted for separately, the local CHF limit is nearly independent of the vessel size. Based upon the scaling law developed in this work, it is possible to merge, within the experimental uncertainties, all the available local CHF data obtained for various vessel sizes under both saturated and subcooled boiling conditions into a single curve. Applications of the scaling law to commercial-size vessels have been made for various system pressures and water levels above the heated vessel. Over the range of conditions explored in this study, the local CHF limit is found to increase by a factor of two or more from the bottom center to the upper edge of the vessel. Meanwhile, the critical heat flux at a given angular position of the heated vessel is also found to increase appreciably with the system pressure and the water level.

  5. Thermal hydraulics characterization of the core and the reactor vessel type BWR; Caracterizacion termohidraulica del nucleo y de la vasija de un reactor tipo BWR

    Energy Technology Data Exchange (ETDEWEB)

    Zapata Y, M.; Lopez H, L.E. [CFE, Carretera Cardel-Nautla Km. 42.5, Municipio Alto Lucero, Veracruz (Mexico)]. e-mail: marxlenin.zapata@cfe.gob.mx

    2008-07-01

    The thermal hydraulics design of a reactor type BWR 5 as the employees in the nuclear power plant of Laguna Verde involves the coupling of at least six control volumes: Pumps jet region, Stratification region, Core region, Vapor dryer region, Humidity separator region and Reactor region. Except by the regions of the core and reactor, these control volumes only are used for design considerations and their importance as operative data source is limited. It is for that is fundamental to complement the thermal hydraulics relations to obtain major data that allow to determine the efficiency of internal components, such as pumps jet, humidity separator and vapor dryer. Like example of the previous thing, calculations are realized on the humidity of the principal vapor during starting, comparing it with the values at the moment incorporated in the data banks of the computers of process of both units. (Author)

  6. Hysteresis in Metal Hydrides.

    Science.gov (United States)

    Flanagan, Ted B., And Others

    1987-01-01

    This paper describes a reproducible process where the irreversibility can be readily evaluated and provides a thermodynamic description of the important phenomenon of hysteresis. A metal hydride is used because hysteresis is observed during the formation and decomposition of the hydride phase. (RH)

  7. 反应堆退役压力容器放射性活度估算方法%Method for Estimation of Activity in Decommissioned Nuclear Reactor Pressure Vessel

    Institute of Scientific and Technical Information of China (English)

    郭武仁; 林晓玲; 郑宁宁

    2011-01-01

    The theoretical calculation and experimental measurement methods for estimation of activity in the decommissioned nuclear reactor pressure vessel were introduced. The physical estimation model was described,and Monte Carlo compute code and 0RIGEN2 code were recommended to be employed for the calculation of the neutron flux and activity in the reactor pressure vessel. Two methods commonly used for determining the activity in the reactor pressure vessel were introduced in detail,I.e.,sampling from the reactor pressure vessel and from the irradiation tube. The neutron flux profile was established to predict the activity in the reactor pressure vessel.%介绍了反应堆退役压力容器放射性活度估算的理论计算和实验测定方法.描述了物理估算模型,推荐采用蒙特卡罗程序和ORIGEN2程序分别计算中子通量密度和放射性活度.对确定压力容器的放射性活度时经常使用的两种方法(压力容器直接取样分析和对辐照监督管取样分析)做了详细介绍.建立了推算压力容器的放射性活度中子通量密度比例曲线.

  8. Review of the International Atomic Energy Agency International database on reactor pressure vessel materials and US Nuclear Regulatory Commission/Oak Ridge National Laboratory embrittlement data base

    Energy Technology Data Exchange (ETDEWEB)

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

    1998-02-01

    The International Atomic Energy Agency (IAEA) has supported neutron radiation effects information exchange through meetings and conferences since the mid-1960s. Through an International Working Group on Reliability of Reactor Pressure Components, information exchange and research activities were fostered through the Coordinated Research Program (CRP) sponsored by the IAEA. The final CRP meeting was held in November 1993, where it was recommended that the IAEA coordinate the development of an International Database on Reactor Pressure Vessel Material (IDRPVM) as the first step in generating an International Database on Aging Management. The purpose of this study was to provide special technical assistance to the NRC in monitoring and evaluating the IAEA activities in developing the IAEA IDRPVM, and to compare the IDRPVM with the Nuclear Regulatory Commission (NRC) - Oak Ridge National Laboratory (ORNL) Power Reactor Embrittlement Data Base (PR-EDB) and provide recommendations for improving the PR-EDB. A first test version of the IDRPVM was distributed at the First Meeting of Liaison Officers to the IAEA IDRPVM, in November 1996. No power reactor surveillance data were included in this version; the testing data were mainly from CRP Phase III data. Therefore, because of insufficient data and a lack of power reactor surveillance data received from the IAEA IDRPVM, the comparison is made based only on the structure of the IDRPVM. In general, the IDRPVM and the EDB have very similar data structure and data format. One anticipates that because the IDRPVM data will be collected from so many different sources, quality assurance of the data will be a difficult task. The consistency of experimental test results will be an important issue. A very wide spectrum of material characteristics of RPV steels and irradiation environments exists among the various countries. Hence the development of embrittlement prediction models will be a formidable task. 4 refs., 2 figs., 4 tabs.

  9. Microstructure and embrittlement of VVER 440 reactor pressure vessel steels; Microstructure et fragilisation des aciers de cuve des reacteurs nucleaires VVER 440

    Energy Technology Data Exchange (ETDEWEB)

    Hennion, A

    1999-03-15

    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)

  10. Effect of thermo-mechanical cycling on zirconium hydride reorientation studied in situ with synchrotron X-ray diffraction

    Science.gov (United States)

    Colas, Kimberly B.; Motta, Arthur T.; Daymond, Mark R.; Almer, Jonathan D.

    2013-09-01

    The circumferential hydrides normally present in nuclear reactor fuel cladding after reactor exposure may dissolve during drying for dry storage and re-precipitate when cooled under load into a more radial orientation, which could embrittle the fuel cladding. It is necessary to study the rates and conditions under which hydride reorientation may happen in order to assess fuel integrity in dry storage. The objective of this work is to study the effect of applied stress and thermal cycling on the hydride morphology in cold-worked stress-relieved Zircaloy-4 by combining conventional metallography and in situ X-ray diffraction techniques. Metallography is used to study the evolution of hydride morphology after several thermo-mechanical cycles. In situ X-ray diffraction performed at the Advanced Photon Source synchrotron provides real-time information on the process of hydride dissolution and precipitation under stress during several thermal cycles. The detailed study of diffracted intensity, peak position and full-width at half-maximum provides information on precipitation kinetics, elastic strains and other characteristics of the hydride precipitation process. The results show that thermo-mechanical cycling significantly increases the radial hydride fraction as well as the hydride length and connectivity. The radial hydrides are observed to precipitate at a lower temperature than circumferential hydrides. Variations in the magnitude and range of hydride strains due to reorientation and cycling have also been observed. These results are discussed in light of existing models and experiments on hydride reorientation. The study of hydride elastic strains during precipitation shows marked differences between circumferential and radial hydrides, which can be used to investigate the reorientation process. Cycling under stress above the threshold stress for reorientation drastically increases both the reoriented hydride fraction and the hydride size. The reoriented hydride

  11. Civacuve analysis software for mis machine examination of pressurized water reactor vessels; Civacuve logiciel d'analyse des controles mis des cuves de reacteurs nucleaires

    Energy Technology Data Exchange (ETDEWEB)

    Dubois, Ph.; Gagnor, A. [Intercontrole, 94 - Rungis (France)

    2001-07-01

    The product software CIVACUVE is used by INTERCONTROLE for the analysis of UT examinations, for detection, performed by the In-Service Inspection Machine (MIS) of the vessels of nuclear power plants. This software is based on an adaptation of an algorithm of SEGMENTATION (CEA CEREM), which is applied prior to any analysis. It is equipped with tools adapted to industrial use. It allows to: - perform image analysis thanks to advanced graphic tools (Zooms, True Bscan, 'contour' selection...), - backup of all data in a database (complete and transparent backup of all informations used and obtained during the different analysis operations), - connect PC to the Database (export of Reports and even of segmented points), - issue Examination Reports, Operating Condition Sheets, Sizing curves... - and last, perform a graphic and numerical comparison between different inspections of the same vessel. Used in Belgium and France on different kind of reactor vessels, CIVACUVE has allowed to show that the principle of SEGMENTATION can be adapted to detection exams. The use of CIVACUVE generates a important time gain as well as the betterment of quality in analysis. Wide data opening toward PC's allows a real flexibility with regard to client's requirements and preoccupations.

  12. Reactor Physics

    Energy Technology Data Exchange (ETDEWEB)

    Ait Abderrahim, A

    2001-04-01

    The Reactor Physics and MYRRHA Department of SCK-CEN offers expertise in various areas of reactor physics, in particular in neutronics calculations, reactor dosimetry, reactor operation, reactor safety and control and non-destructive analysis of reactor fuel. This expertise is applied in the Department's own research projects in the VENUS critical facility, in the BR1 reactor and in the MYRRHA project (this project aims at designing a prototype Accelerator Driven System). Available expertise is also used in programmes external to the Department such as the reactor pressure steel vessel programme, the BR2 reactor dosimetry, and the preparation and interpretation of irradiation experiments by means of neutron and gamma calculations. The activities of the Fuzzy Logic and Intelligent Technologies in Nuclear Science programme cover several domains outside the department. Progress and achievements in these topical areas in 2000 are summarised.

  13. Reactor

    Science.gov (United States)

    Evans, Robert M.

    1976-10-05

    1. A neutronic reactor having a moderator, coolant tubes traversing the moderator from an inlet end to an outlet end, bodies of material fissionable by neutrons of thermal energy disposed within the coolant tubes, and means for circulating water through said coolant tubes characterized by the improved construction wherein the coolant tubes are constructed of aluminum having an outer diameter of 1.729 inches and a wall thickness of 0.059 inch, and the means for circulating a liquid coolant through the tubes includes a source of water at a pressure of approximately 350 pounds per square inch connected to the inlet end of the tubes, and said construction including a pressure reducing orifice disposed at the inlet ends of the tubes reducing the pressure of the water by approximately 150 pounds per square inch.

  14. The Effective Convectivity Model for Simulation and Analysis of Melt Pool Heat Transfer in a Light Water Reactor Pressure Vessel Lower Head

    Energy Technology Data Exchange (ETDEWEB)

    Tran, Chi Thanh

    2009-09-15

    Severe accidents in a Light Water Reactor (LWR) have been a subject of intense research for the last three decades. The research in this area aims to reach understanding of the inherent physical phenomena and reduce the uncertainties in their quantification, with the ultimate goal of developing models that can be applied to safety analysis of nuclear reactors, and to evaluation of the proposed accident management schemes for mitigating the consequences of severe accidents. In a hypothetical severe accident there is likelihood that the core materials will be relocated to the lower plenum and form a decay-heated debris bed (debris cake) or a melt pool. Interactions of core debris or melt with the reactor structures depend to a large extent on the debris bed or melt pool thermal hydraulics. In case of inadequate cooling, the excessive heat would drive the structures' overheating and ablation, and hence govern the vessel failure mode and timing. In turn, threats to containment integrity associated with potential ex-vessel steam explosions and ex-vessel debris uncoolability depend on the composition, superheat, and amount of molten corium available for discharge upon the vessel failure. That is why predictions of transient melt pool heat transfer in the reactor lower head, subsequent vessel failure modes and melt characteristics upon the discharge are of paramount importance for plant safety assessment. The main purpose of the present study is to develop a method for reliable prediction of melt pool thermal hydraulics, namely to establish a computational platform for cost-effective, sufficiently-accurate numerical simulations and analyses of core Melt-Structure-Water Interactions in the LWR lower head during a postulated severe core-melting accident. To achieve the goal, an approach to efficient use of Computational Fluid Dynamics (CFD) has been proposed to guide and support the development of models suitable for accident analysis. The CFD method, on the one hand

  15. J-integral elastic plastic fracture mechanics evaluation of the stability of cracks in nuclear reactor pressure vessels

    Energy Technology Data Exchange (ETDEWEB)

    Gomez, M. P.; McMeeking, R. M.; Parks, D. M.

    1980-06-01

    Contributions were made toward developing a new methodology to assess the stability of cracks in pressure vessels made from materials that exhibit a significant increase in toughness during the early increments of crack growth. It has a wide range of validity from linear elastic to fully plastic behavior.

  16. Regenerative Hydride Heat Pump

    Science.gov (United States)

    Jones, Jack A.

    1992-01-01

    Hydride heat pump features regenerative heating and single circulation loop. Counterflow heat exchangers accommodate different temperatures of FeTi and LaNi4.7Al0.3 subloops. Heating scheme increases efficiency.

  17. Regenerative Hydride Heat Pump

    Science.gov (United States)

    Jones, Jack A.

    1992-01-01

    Hydride heat pump features regenerative heating and single circulation loop. Counterflow heat exchangers accommodate different temperatures of FeTi and LaNi4.7Al0.3 subloops. Heating scheme increases efficiency.

  18. Contributions of Cu-rich clusters, dislocation loops and nanovoids to the irradiation-induced hardening of Cu-bearing low-Ni reactor pressure vessel steels

    Science.gov (United States)

    Bergner, F.; Gillemot, F.; Hernández-Mayoral, M.; Serrano, M.; Török, G.; Ulbricht, A.; Altstadt, E.

    2015-06-01

    Dislocation loops, nanovoids and Cu-rich clusters (CRPs) are known to represent obstacles for dislocation glide in neutron-irradiated reactor pressure vessel (RPV) steels, but a consistent experimental determination of the respective obstacle strengths is still missing. A set of Cu-bearing low-Ni RPV steels and model alloys was characterized by means of SANS and TEM in order to specify mean size and number density of loops, nanovoids and CRPs. The obstacle strengths of these families were estimated by solving an over-determined set of linear equations. We have found that nanovoids are stronger than loops and loops are stronger than CRPs. Nevertheless, CRPs contribute most to irradiation hardening because of their high number density. Nanovoids were only observed for neutron fluences beyond typical end-of-life conditions of RPVs. The estimates of the obstacle strength are critically compared with reported literature data.

  19. Experimental and numerical investigation of coolant mixing in a model of reactor pressure vessel down-comer and in cold leg inlets

    Directory of Open Access Journals (Sweden)

    Hutli Ezddin

    2017-01-01

    Full Text Available Thermal fatigue and pressurized thermal shock phenomena are the main problems for the reactor pressure vessel and the T-junctions both of them depend on the mixing of the coolant. The mixing process, flow and temperature distribution has been investigated experimentally using particle image velocimetry, laser induced fluorescence, and simulated by CFD tools. The obtained results showed that the ratio of flow rate between the main pipe and the branch pipe has a big influence on the mixing process. The particle image velocimetry/planar laser-induced fluorescence measurements technologies proved to be suitable for the investigation of turbulent mixing in the complicated flow system: both velocity and temperature distribution are important parameters in the determination of thermal fatigue and pressurized thermal shock. Results of the applied these techniques showed that both of them can be used as a good provider for data base and to validate CFD results.

  20. Study of atomic clusters in neutron irradiated reactor pressure vessel surveillance samples by extended X-ray absorption fine structure spectroscopy

    Energy Technology Data Exchange (ETDEWEB)

    Cammelli, S. [LWV, NES, Paul Scherrer Institute, 5232 Villigen PSI (Switzerland); Fachbereich C - Physik, Bergische Universitaet Wuppertal, Gauss-Str. 20, 42097 Wuppertal (Germany)], E-mail: Sebastiano.cammelli@psi.ch; Degueldre, C.; Kuri, G.; Bertsch, J. [LWV, NES, Paul Scherrer Institute, 5232 Villigen PSI (Switzerland); Luetzenkirchen-Hecht, D.; Frahm, R. [Fachbereich C - Physik, Bergische Universitaet Wuppertal, Gauss-Str. 20, 42097 Wuppertal (Germany)

    2009-03-31

    Copper and nickel impurities in nuclear reactor pressure vessel (RPV) steel can form nano-clusters, which have a strong impact on the ductile-brittle transition temperature of the material. Thus, for control purposes and simulation of long irradiation times, surveillance samples are submitted to enhanced neutron irradiation. In this work, surveillance samples from a Swiss nuclear power plant were investigated by extended X-ray absorption fine structure spectroscopy (EXAFS). The density of Cu and Ni atoms determined in the first and second shells around the absorber is affected by the irradiation and temperature. The comparison of the EXAFS data at Cu and Ni K-edges shows that these elements reside in arrangements similar to bcc Fe. However, the EXAFS analysis reveals local irradiation damage in the form of vacancy fractions, which can be determined with a precision of {approx}5%. There are indications that the formation of Cu and Ni clusters differs significantly.

  1. Fracture mechanics characterisation of the beltline welding seam of the decommissioned WWER-440 reactor pressure vessel of nuclear power plant Greifswald Unit 4

    Energy Technology Data Exchange (ETDEWEB)

    Viehrig, Hans-Werner, E-mail: H.W.Viehrig@hzdr.de [Helmholz Zentrum Dresden-Rossendorf, P.O. Box 510119, 01314 Dresden (Germany); Altstadt, Eberhard; Houska, Mario [Helmholz Zentrum Dresden-Rossendorf, P.O. Box 510119, 01314 Dresden (Germany); Valo, Matti [VTT Manufacturing Technology, P.O. Box 17042, 02044 VTT (Finland)

    2012-01-15

    The paper presents data measured for trepans sampled from the decommissioned WWER-440 reactor pressure vessel of the NPP Greifswald Unit 4. The main focus of this work is on fracture toughness characterisation according to test standard ASTM E1921. Large variations in the evaluated reference temperature values, T{sub 0}, across the wall of the multilayer beltline welding seam were observed. Generally, the ductile-to-brittle transition temperature shift predicted by the Russian code for the present content of deleterious elements P and Cu and the accumulated neutron fluences lies within the amount of the scatter of the measured T{sub 0} values. Metallographic investigations show that the T{sub 0} values measured with T-S oriented Charpy size SE(B) specimens from different thickness locations of the multilayer welding seams strongly depend on the microstructure at the specimen crack tip, and, consequently, on the initial structure of the multilayer welding seam. The RPV integrity is discussed, taking into account a pressurised thermal shock scenario. - Highlights: Black-Right-Pointing-Pointer The paper presents data of samples from a decommissioned reactor pressure vessel. Black-Right-Pointing-Pointer The main focus is on fracture toughness characterisation of the beltline weld seam. Black-Right-Pointing-Pointer Large variation in the evaluated reference temperatures T{sub 0} was observed. Black-Right-Pointing-Pointer T{sub 0} values strongly depend on the microstructure at the specimen crack tip. Black-Right-Pointing-Pointer RPV integrity is discussed, taking into account a pressurised thermal shock scenario.

  2. The Assessment and Validation of Mini-Compact Tension Test Specimen Geometry and Progress in Establishing Technique for Fracture Toughness Master Curves for Reactor Pressure Vessel Steels

    Energy Technology Data Exchange (ETDEWEB)

    Sokolov, Mikhail A. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Nanstad, Randy K. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2016-09-01

    Small specimens are playing the key role in evaluating properties of irradiated materials. The use of small specimens provides several advantages. Typically, only a small volume of material can be irradiated in a reactor at desirable conditions in terms of temperature, neutron flux, and neutron dose. A small volume of irradiated material may also allow for easier handling of specimens. Smaller specimens reduce the amount of radioactive material, minimizing personnel exposures and waste disposal. However, use of small specimens imposes a variety of challenges as well. These challenges are associated with proper accounting for size effects and transferability of small specimen data to the real structures of interest. Any fracture toughness specimen that can be made out of the broken halves of standard Charpy specimens may have exceptional utility for evaluation of reactor pressure vessels (RPVs) since it would allow one to determine and monitor directly actual fracture toughness instead of requiring indirect predictions using correlations established with impact data. The Charpy V-notch specimen is the most commonly used specimen geometry in surveillance programs. Assessment and validation of mini-CT specimen geometry has been performed on previously well characterized HSST Plate 13B, an A533B class 1 steel. It was shown that the fracture toughness transition temperature measured by these Mini-CT specimens is within the range of To values that were derived from various large fracture toughness specimens. Moreover, the scatter of the fracture toughness values measured by Mini-CT specimens perfectly follows the Weibull distribution function providing additional proof for validation of this geometry for the Master Curve evaluation of rector pressure vessel steels. Moreover, the International collaborative program has been developed to extend the assessment and validation efforts to irradiated weld metal. The program is underway and involves ORNL, CRIEPI, and EPRI.

  3. A Review of the Application of Rate Theory to Simulate Vacancy Cluster Formation and Interstitial Defect Formation in Reactor Pressure Vessel Steel

    Directory of Open Access Journals (Sweden)

    Fallon Laliberte

    2015-10-01

    Full Text Available The beltline region of the reactor pressure vessel (RPV is subject to an extreme radiation, temperature, and pressure environment over several decades of operation; therefore it is necessary to understand the mechanisms through which radiation damage occurs and how it affects the mechanical and chemical properties of the RPV steel. Chemical rate theory is a mean field rate theory simulation model which applies chemistry to the evaluation of irradiation-induced embrittlement. It presents one method of analysis that may be coupled to other distinct methods, in order to analyze defect formation, ultimately providing useful information on strength, ductility, toughness and dimensional stability changes for effects such as embrittlement, reduction in ductility and toughness, void swelling, hardening, irradiation creep, stress corrosion cracking, etc. over time as materials are subjected to reactor operational irradiation. This paper serves as a brief review of rate theory fundamentals and presents several examples of research that exemplify the application and importance of rate theory in examining the effects of radiation damage on RPV steel.

  4. Experimental study on buoyancy-driven exchange flows through breaches of a tokamak vacuum vessel in a fusion reactor under the loss-of-vacuum-event conditions

    Energy Technology Data Exchange (ETDEWEB)

    Takase, Kazuyuki; Tomoaki, Kunugi; Ogawa, Masurou; Seki, Yasushi [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan)

    1997-02-01

    As one of thermofluid safety studies in the International Thermonuclear Experimental Reactor, buoyancy-driven exchange flow behavior through breaches of a vacuum vessel (VV) has been investigated quantitatively by using a preliminary loss-of-vacuum-event (LOVA) apparatus that simulated the tokamak VV of a fusion reactor with a small-scaled model. To carry out the present experiments under the atmospheric pressure condition, helium gas and air were provided as the working fluids. The inside of the VV was initially filled with helium gas and the outside was atmosphere. The breaches on the VV under the LOVA condition were simulated by opening six simulated breaches to which were set the different positions on the VV. When the buoyancy-driven exchange flow through the breach occurred, helium gas went out from the inside of the VV through the breach to the outside and air flowed into the inside of the VV through the breach from the outside. The exchange rate in the VV between helium gas and air was calculated from the measured weight change of the VV with time since the experiment has started. experimental parameters were breach position, breach number, breach length, breach size, and breach combination. The present study clarifies that the relation between the exchange rate and the breach position of the VV depended on the magnitude of the potential energy from the ground level to the breach position, and then, the exchange rate decreased as the breach length increased and as the breach size decreased.

  5. Influence of Modelling Options in RELAP5/SCDAPSIM and MAAP4 Computer Codes on Core Melt Progression and Reactor Pressure Vessel Integrity

    Directory of Open Access Journals (Sweden)

    Siniša Šadek

    2010-01-01

    Full Text Available RELAP5/SCDAPSIM and MAAP4 are two widely used severe accident computer codes for the integral analysis of the core and the reactor pressure vessel behaviour following the core degradation. The objective of the paper is the comparison of code results obtained by application of different modelling options and the evaluation of influence of thermal hydraulic behaviour of the plant on core damage progression. The analysed transient was postulated station blackout in NPP Krško with a leakage from reactor coolant pump seals. Two groups of calculations were performed where each group had a different break area and, thus, a different leakage rate. Analyses have shown that MAAP4 results were more sensitive to varying thermal hydraulic conditions in the primary system. User-defined parameters had to be carefully selected when the MAAP4 model was developed, in contrast to the RELAP5/SCDAPSIM model where those parameters did not have any significant impact on final results.

  6. Hydriding performances and modeling of a small-scale ZrCo bed

    Energy Technology Data Exchange (ETDEWEB)

    Koo, D.; Lee, J.; Park, J.; Paek, S.; Chung, H. [KAERI-UST, Yuseong, Daejeon (Korea, Republic of); Chang, M.H.; Yun, S.H.; Cho, S.; Jung, K.J. [NFRI, Yuseong, Daejeon (Korea, Republic of)

    2015-03-15

    In order to evaluate the performance of the hydriding of a ZrCo bed, a small-scale getter bed of ZrCo was designed and fabricated. The results show that the hydriding time at room temperature was somewhat shorter than that at higher temperatures of ZrCo and that the performance of hydriding at low temperatures of ZrCo was better than that at high temperatures of ZrCo. The experimental results of the hydrogen pressure of hydriding (ZrCoH{sub 2.8}) at different temperatures were in agreement with the computed values using a numerical modeling equation but with a small difference during the first 10 minutes of the hydriding of ZrCo. The model is based on the Kozeny-Carman equation. The effect of a helium blanket on hydriding was measured and analyzed. The hydriding with no helium blanket in the primary vessel of ZrCo is much faster than that with a helium blanket. The hydriding at a helium concentration of 8% is slower than that at 0%. As the helium concentration increases, the hydriding of ZrCo decreases. The experimental results of the hydriding with 0 %, 4%, and 8% of helium concentration are in agreement with the calculated values but with minimal differences during the first 10 minutes.

  7. Towards the prediction of the rupture of a pressurized water reactor vessel in case of accident; Vers la prevision du dechirement d'une cuve de reacteur a eau pressurisee en cas d'accident

    Energy Technology Data Exchange (ETDEWEB)

    Tardif, N.; Coret, M.; Combescure, A. [Lyon Univ., CNRS, INSA-Lyon, LaMCoS UMR5259, 69 (France); Tardif, N.; Nicaise, G. [Institut de Radioprotection et de Surete Nucleaire, DSR/SAGR/BPhAG, 92 - Fontenay-aux-Roses (France)

    2009-07-01

    Through a scale model of a reactor vessel submitted to a thermal and mechanical load during a severe accident, it is possible to follow the initiation and propagation of cracks in real time by tests carried out on laboratory. (O.M.)

  8. Lightweight hydride storage materials

    Energy Technology Data Exchange (ETDEWEB)

    Thomas, G.J.; Guthrie, S.E.; Bauer, W. [Sandia National Labs., Livermore, CA (United States)

    1995-09-01

    The need for lightweight hydrides in vehicular applications has prompted considerable research into the use of magnesium and its alloys. Although this earlier work has provided some improved performance in operating temperature and pressure, substantial improvements are needed before these materials will significantly enhance the performance of an engineered system on a vehicle. We are extending the work of previous investigators on Mg alloys to reduce the operating temperature and hydride heat of formation in light weight materials. Two important results will be discussed in this paper: (1) a promising new alloy hydride was found which has better pressure-temperature characteristics than any previous Mg alloy and, (2) a new fabrication process for existing Mg alloys was developed and demonstrated. The new alloy hydride is composed of magnesium, aluminum and nickel. It has an equilibrium hydrogen overpressure of 1.3 atm. at 200{degrees}C and a storage capacity between 3 and 4 wt.% hydrogen. A hydrogen release rate of approximately 5 x 10{sup -4} moles-H{sub 2}/gm-min was measured at 200{degrees}C. The hydride heat of formation was found to be 13.5 - 14 kcal/mole-H{sub 2}, somewhat lower than Mg{sub 2}Ni. The new fabrication method takes advantage of the high vapor transport of magnesium. It was found that Mg{sub 2}Ni produced by our low temperature process was better than conventional materials because it was single phase (no Mg phase) and could be fabricated with very small particle sizes. Hydride measurements on this material showed faster kinetic response than conventional material. The technique could potentially be applied to in-situ hydride bed fabrication with improved packing density, release kinetics, thermal properties and mechanical stability.

  9. Master curve analysis of the SA508 Gr. 4N Ni-Mo-Cr low alloy steels for reactor pressure vessels

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Ki Hyoung; Wee, Dang Moon [KAIST, Daejeon (Korea, Republic of); Kim, Min Chul; Lee, Bong Sang [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2009-10-15

    Low alloy steels used as Reactor Pressure Vessels (RPVs) materials directly relate to the safety margin and the life span of reactors. Currently, SA508 Gr.3 low alloy steel is generally used for RPV material. But, for larger capacity and long-term durability of RPV, materials that have better properties including strength and toughness are needed. Therefore, tempered martensitic SA508 Gr.4N low alloy steel is considered as a candidate material due to excellent mechanical properties. The fracture toughness loss caused by irradiation embrittlement during reactor operation is one of the important issues for ferritic RPV steels, because the decrease of fracture toughness is directly related to the integrity of RPVs. One reliable and efficient concept to evaluate the fracture toughness of ferritic steels is master curve method. In ASTM E1921, it is clearly mentioned the universal shape of the median toughness-temperature curve for ferritic steels including tempered martensitic steels. However, currently, concerns have arisen regarding the appropriateness of the universal shape in ASTM for the tempered martensitic steels such as Eurofer97. Therefore, it may be necessary to assess the master curve applicability for the tempered martensitic SA508 Gr.4N low alloy steel. In this study, the fracture toughness behavior with temperature of the tempered martensitic SA508 Gr.4N low alloy steels was evaluated using the ASTM E1921 master curve method. And the results were compared with those of the bainitic SA508 Gr.3 low alloy steel. Furthermore, the way to define the fracture toughness behavior of Gr.4N steels well is discussed.

  10. Study on operation conditions and an operation system of a nuclear powered submersible research vessel, 'report of working group on application of a very small nuclear reactor to an ocean research'

    Energy Technology Data Exchange (ETDEWEB)

    Ura, Tamaki [Tokyo Univ., Tokyo (Japan); Takamasa, Tomoji [Tokyo Univ. of Mercantile Marine, Tokyo (Japan); Nishimura, Hajime [Japan Marine Science and Technology Center, Yokosuka, Kanagawa (JP)] [and others

    2001-07-01

    JAERI has studied on design of a nuclear powered submersible research vessel, which will navigate under sea mainly in the Arctic Ocean, as a part of the design activity of advanced marine reactors. This report describes operation conditions and an operating system of the vessel, which were discussed by the specialists of hull design, sound positioning, ship motions and oceanography, etc. The design conditions on ship motions for submersible vessels were surveyed considering regulations in our country, and ship motions were evaluated in the cases of underwater and surface navigations taking account of observation activities in the Arctic Ocean. The effect of ship motions on the compact nuclear reactor SCR was assessed. A submarine transponder system and an on-ice communication buoy system were examined as a positioning and communication system, supposing the activity under ice. The interval between transponders or communication buoys was recommended as 130 km. Procedures to secure safety of nuclear powered submersible research vessel were discussed according to accidents on the hull or the nuclear reactor. These results were reflected to the concept of the nuclear powered submersible research vessel, and subjects to be settled in the next step were clarified. (author)

  11. Leakage Tests of the Stainless Steel Vessels of the Antineutrino Detectors in the Daya Bay Reactor Neutrino Experiment

    CERN Document Server

    Chen, Xiaohui; Heng, Yuekun; Wang, Lingshu; Tang, Xiao; Ma, Xiaoyan; Zhuang, Honglin; Band, Henry; Cherwinka, Jeff; Xiao, Qiang; Heeger, Karsten M

    2012-01-01

    The antineutrino detectors in the Daya Bay reactor neutrino experiment are liquid scintillator detectors designed to detect low energy particles from antineutrino interactions with high efficiency and low backgrounds. Since the antineutrino detector will be installed in a water Cherenkov cosmic ray veto detector and will run for 3 to 5 years, ensuring water tightness is critical to the successful operation of the antineutrino detectors. We choose a special method to seal the detector. Three leak checking methods have been employed to ensure the seal quality. This paper will describe the sealing method and leak testing results.

  12. Analysis and Verification of Direct Vessel Injection Line Break event tree for AP1000 reactor with TRACE code

    Energy Technology Data Exchange (ETDEWEB)

    Queral, C.; Montero-Mayorga, J.; Gonzalez-Cadelo, J.

    2013-07-01

    The AP1000 PRA thermal hydraulic simulations were performed with MAAP code, which allows simulating sequences with low computational efforts. On the other hand, the use of best estimate codes allows verifying PRA results as well as obtaining a greater knowledge of the phenomenology of such sequences. The initiating event with the greatest contribution to core damage is Direct Vessel Injection Line Break (DVILB). This paper presents a review of DVILB sequences of AP1000 with TRACE code for verifying sequences previously analyzed by Westinghouse with MAAP code. The sequences which configure the DVILB event tree during short term have been simulated. The results obtained confirm the ones obtained in AP1000 PRA.

  13. Environmentally-assisted cracking behaviour in the transition region of alloy 182/low-alloy reactor pressure vessel steel dissimilar metal weld joints in simulated boiling water reactor normal water chemistry environment

    Energy Technology Data Exchange (ETDEWEB)

    Seifert, H.P.; Ritter, S.; Leber, H.J. [Paul Scherrer Institute (Switzerland). Lab. for Nuclear Materials

    2010-07-01

    The stress corrosion cracking (SCC) behaviour perpendicular to the fusion line in the transition region between the Alloy 182 nickel-base weld metal and the adjacent low-alloy reactor pressure vessel (RPV) steel of simulated dissimilar metal weld joints was investigated under boiling water reactor normal water chemistry conditions at different stress intensities and chloride concentrations. A special emphasis was placed to the question whether a fast growing interdendritic SCC crack in the highly susceptible Alloy 182 weld metal can easily cross the fusion line and significantly propagate into the adjacent low-alloy RPV steel. Cessation of interdendritic stress corrosion crack growth was observed in high-purity or sulphate-containing oxygenated water under periodical partial unloading or constant loading conditions with stress intensity factors below 60 MPa.m{sup 1/2} for those parts of the crack front, which reached the fusion line. In chloride containing water, on the other hand, the interdendritic stress corrosion crack in the Alloy 182 weld metal very easily crossed the fusion line and further propagated with a very high growth rate as a transgranular crack into the heat-affected zone and base material of the adjacent low-alloy steel. (orig.)

  14. 浅析核电厂反应堆压力容器完整性问题%Discussion about Integrity Issue of Reactor Pressure Vessel in Nuclear Power Plant

    Institute of Scientific and Technical Information of China (English)

    张加军; 陈晶晶; 车树伟; 吴彦农

    2014-01-01

    Nuclear reactor pressure vessel ( RPV) is one of the critical equipments for a nuclear power plant ,and its integrity will impact the safe operation of the nuclear power plant .To ensure the integrity of the reactor pressure vessel ,safety concern is highlighted in design ,manufacture ,installation and operation phase .It provided a reference for the nuclear power plant reactor pressure vessel in design ,manufacture , installation,operation and maintenance phase by introducing the material development of reactor pressure vessel ,typical degradation model of reactor pressure vessel ,and analyzing reasons for the degradation and proposing relevant measures to ensure the integrity of RPV .%反应堆压力容器( RPV)作为核电厂重要主设备之一,其完整性直接影响到核电厂的安全运行,为了确保反应堆压力容器的完整性,需要在设计、制造、安装和运行过程中重点关注相关问题。介绍了反应堆压力容器的材质发展过程、反应堆压力容器的典型降级模式,并对产生降级的原因进行了分析,提出了下一步预防降级可采取的措施,以确保反应堆压力容器的完整性,进而为核电厂的反应堆压力容器的设计、制造、安装和运行维护阶段提供参考。

  15. Development of a Versatile Ultrasonic Internal Pipe/Vessel Component Monitor for In-Service Inspection of Nuclear Reactor Components

    Energy Technology Data Exchange (ETDEWEB)

    Searfass, Clifford T. [Structural Integrity Associates, Inc., State College, PA (United States); Malinowski, Owen M. [Structural Integrity Associates, Inc., State College, PA (United States); Van Velsor, Jason K. [Structural Integrity Associates, Inc., State College, PA (United States)

    2015-03-22

    The stated goal of this work was to develop a versatile system which could accurately measure vessel and valve internal vibrations and cavitation formation under in-service conditions in nuclear power plants, ultrasonically. The developed technology will benefit the nuclear power generation industry by allowing plant operators to monitor valve and vessel internals during operation. This will help reduce planned outages and plant component failures. During the course of this work, Structural Integrity Associates, Inc. gathered information from industry experts that target vibration amplitudes to be detected should be in the range of 0.001-in to 0.005-in (0.025-mm to 0.127-mm) and target vibration frequency ranges which should be detected were found to be between 0-Hz and 300-Hz. During the performed work, an ultrasonic measuring system was developed which utilized ultrasonic pulse-echo time-of-flight measurements to measure vibration frequency and amplitude. The developed system has been shown to be able to measure vibration amplitudes as low as 0.0008-in (0.020-mm) with vibration frequencies in the range of 17-Hz to 1000-Hz. Therefore, the developed system was able to meet the industry needs for vibration measurement. The developed ultrasonic system was also to be able to measure cavitation formation by monitoring the received ultrasonic time- and frequency-domain signals. This work also demonstrated the survivability of commercially available probes at temperatures up to 300-F for several weeks.

  16. An emerging reactor technology for chemical synthesis: surface acoustic wave-assisted closed-vessel Suzuki coupling reactions.

    Science.gov (United States)

    Kulkarni, Ketav; Friend, James; Yeo, Leslie; Perlmutter, Patrick

    2014-07-01

    In this paper we demonstrate the use of an energy-efficient surface acoustic wave (SAW) device for driving closed-vessel SAW-assisted (CVSAW), ligand-free Suzuki couplings in aqueous media. The reactions were carried out on a mmolar scale with low to ultra-low catalyst loadings. The reactions were driven by heating resulting from the penetration of acoustic energy derived from RF Raleigh waves generated by a piezoelectric chip via a renewable fluid coupling layer. The yields were uniformly high and the reactions could be executed without added ligand and in water. In terms of energy density this new technology was determined to be roughly as efficient as microwaves and superior to ultrasound.

  17. Hydrogen, lithium, and lithium hydride production

    Energy Technology Data Exchange (ETDEWEB)

    Brown, Sam W.; Spencer, Larry S.; Phillips, Michael R.; Powell, G. Louis; Campbell, Peggy J.

    2017-06-20

    A method is provided for extracting hydrogen from lithium hydride. The method includes (a) heating lithium hydride to form liquid-phase lithium hydride; (b) extracting hydrogen from the liquid-phase lithium hydride, leaving residual liquid-phase lithium metal; (c) hydriding the residual liquid-phase lithium metal to form refined lithium hydride; and repeating steps (a) and (b) on the refined lithium hydride.

  18. Effect of CTE on Fatigue Cracking of Stainless Steel Vessels

    Energy Technology Data Exchange (ETDEWEB)

    Bird, E. L.; Mustaleski, T. M.

    2002-01-31

    Visual examination of lithium hydride reactor vessels revealed cracks that were adjacent to welds. Most cracks were parallel to the weld in the bottom portion of the vessel. Sections were cut out of the vessel containing these cracks and examined using the metallograph, scanning electron microscope, and microprobe to determine the cause of cracking. most of the cracks originated on the outer surface just outside the weld fusion line in the heat affected zone and propagated along grain boundaries. Crack depth of those sections examined ranged from about 300 to 500 {micro}m. Other cracks were reported to have reached a maximum depth of 0.32-cm (0.125-inch). The primary cause of cracking was the creation of high tensile stresses associated with the CTE differences between the filler metal and the base metal during operation of the vessel in a thermally cyclic environment. This failure mechanism could be described as creep-type fatigue whereby crack propagation might have been aided by the presence of brittle chromium carbides along the grain boundaries, which is indicative of a slightly sensitized microstructure.

  19. Lithium hydride - A space age shielding material

    Science.gov (United States)

    Welch, F. H.

    1974-01-01

    Men and materials performing in the environment of an operating nuclear reactor require shielding from the escaping neutron particles and gamma rays. For efficient shielding from gamma rays, dense, high atomic number elements such as iron, lead, or tungsten are required, whereas light, low atomic number elements such as hydrogen, lithium, or beryllium are required for efficient neutron shielding. The use of lithium hydride (LiH) as a highly efficient neutron-shielding material is considered. It contains, combined into a single, stable compound, two of the elements most effective in attenuating and absorbing neutrons.

  20. Research on friction coefficient of nuclear Reactor Vessel Internals Hold Down Spring: Stress coefficient test analysis method

    Energy Technology Data Exchange (ETDEWEB)

    Linjun, Xie, E-mail: linjunx@zjut.edu.cn [College of Mechanical Engineering, Zhejiang University of Technology, Hangzhou 310014 (China); Guohong, Xue; Ming, Zhang [Shanghai Nuclear Engineering Research & Design Institute, Shanghai 200233 (China)

    2016-08-01

    Graphical abstract: HDS stress coefficient test apparatus. - Highlights: • This paper performs mathematic deduction to the physical model of Hold Down Spring (HDS), establishes a mathematic model of axial load P and stress, stress coefficient and friction coefficient and designs a set of test apparatuses for simulating the pretightening process of the HDS for the first time according to a model similarity criterion. • The mathematical relation between the load and the strain is obtained about the HDS, and the mathematical model of the stress coefficient and the friction coefficient is established. So, a set of test apparatuses for obtaining the stress coefficient is designed according to the model scaling criterion and the friction coefficient of the K1000 HDS is calculated to be 0.336 through the obtained stress coefficient. • The relation curve between the theoretical load and the friction coefficient is obtained through analysis and indicates that the change of the friction coefficient f would influence the pretightening load under the condition of designed stress. The necessary pretightening load in the design process is calculated to be 5469 kN according to the obtained friction coefficient. Therefore, the friction coefficient and the pretightening load under the design conditions can provide accurate pretightening data for the analysis and design of the reactor HDS according to the operations. - Abstract: This paper performs mathematic deduction to the physical model of Hold Down Spring (HDS), establishes a mathematic model of axial load P and stress, stress coefficient and friction coefficient and designs a set of test apparatuses for simulating the pretightening process of the HDS for the first time according to a model similarity criterion. By carrying out tests and researches through a stress testing technique, P–σ curves in loading and unloading processes of the HDS are obtained and the stress coefficient k{sub f} of the HDS is obtained. So, the

  1. PERFORM 60 - Prediction of the effects of radiation for reactor pressure vessel and in-core materials using multi-scale modelling - 60 years foreseen plant lifetime

    Energy Technology Data Exchange (ETDEWEB)

    Leclercq, Sylvain, E-mail: sylvain.leclercq@edf.f [EDF R and D, Materials and Mechanics of Components, Avenue des Renardieres - Ecuelles, 77818 Moret sur Loing Cedex (France); Lidbury, David [SERCO Assurance - Walton House, 404 Faraday Street, Birchwood Park, Warrington, Cheshire WA3 6GA (United Kingdom); Van Dyck, Steven [SCK-CEN, Nuclear Material Science, Boeretang 200, BE, 2400 Mol (Belgium); Moinereau, Dominique [EDF R and D, Materials and Mechanics of Components, Avenue des Renardieres - Ecuelles, 77818 Moret sur Loing Cedex (France); Alamo, Ana [CEA Saclay, DEN/DSOE, 91191 Gif-sur-Yvette (France); Mazouzi, Abdou Al [EDF R and D, Materials and Mechanics of Components, Avenue des Renardieres - Ecuelles, 77818 Moret sur Loing Cedex (France)

    2010-11-01

    In nuclear power plants, materials may undergo degradation due to severe irradiation conditions that may limit their operational life. Utilities that operate these reactors need to quantify the ageing and the potential degradations of some essential structures of the power plant to ensure safe and reliable plant operation. So far, the material databases needed to take account of these degradations in the design and safe operation of installations mainly rely on long-term irradiation programs in test reactors as well as on mechanical or corrosion testing in specialized hot cells. Continuous progress in the physical understanding of the phenomena involved in irradiation damage and continuous progress in computer sciences have now made possible the development of multi-scale numerical tools able to simulate the effects of irradiation on materials microstructure. A first step towards this goal has been successfully reached through the development of the RPV-2 and Toughness Module numerical tools by the scientific community created around the FP6 PERFECT project. These tools allow to simulate irradiation effects on the constitutive behaviour of the reactor pressure vessel low alloy steel, and also on its failure properties. Relying on the existing PERFECT Roadmap, the 4 years Collaborative Project PERFORM 60 has mainly for objective to develop multi-scale tools aimed at predicting the combined effects of irradiation and corrosion on internals (austenitic stainless steels) and also to improve existing ones on RPV (bainitic steels). PERFORM 60 is based on two technical sub-projects: (i) RPV and (ii) internals. In addition to these technical sub-projects, the Users' Group and Training sub-project shall allow representatives of constructors, utilities, research organizations... from Europe, USA and Japan to receive the information and training to get their own appraisal on limits and potentialities of the developed tools. An important effort will also be made to teach

  2. PERFORM 60 - Prediction of the effects of radiation for reactor pressure vessel and in-core materials using multi-scale modelling - 60 years foreseen plant lifetime

    Science.gov (United States)

    Leclercq, Sylvain; Lidbury, David; Van Dyck, Steven; Moinereau, Dominique; Alamo, Ana; Mazouzi, Abdou Al

    2010-11-01

    In nuclear power plants, materials may undergo degradation due to severe irradiation conditions that may limit their operational life. Utilities that operate these reactors need to quantify the ageing and the potential degradations of some essential structures of the power plant to ensure safe and reliable plant operation. So far, the material databases needed to take account of these degradations in the design and safe operation of installations mainly rely on long-term irradiation programs in test reactors as well as on mechanical or corrosion testing in specialized hot cells. Continuous progress in the physical understanding of the phenomena involved in irradiation damage and continuous progress in computer sciences have now made possible the development of multi-scale numerical tools able to simulate the effects of irradiation on materials microstructure. A first step towards this goal has been successfully reached through the development of the RPV-2 and Toughness Module numerical tools by the scientific community created around the FP6 PERFECT project. These tools allow to simulate irradiation effects on the constitutive behaviour of the reactor pressure vessel low alloy steel, and also on its failure properties. Relying on the existing PERFECT Roadmap, the 4 years Collaborative Project PERFORM 60 has mainly for objective to develop multi-scale tools aimed at predicting the combined effects of irradiation and corrosion on internals (austenitic stainless steels) and also to improve existing ones on RPV (bainitic steels). PERFORM 60 is based on two technical sub-projects: (i) RPV and (ii) internals. In addition to these technical sub-projects, the Users' Group and Training sub-project shall allow representatives of constructors, utilities, research organizations… from Europe, USA and Japan to receive the information and training to get their own appraisal on limits and potentialities of the developed tools. An important effort will also be made to teach young

  3. Modeling of Gallium Nitride Hydride Vapor Phase Epitaxy

    Science.gov (United States)

    Meyyappan, Meyya; Arnold, James O. (Technical Monitor)

    1997-01-01

    A reactor model for the hydride vapor phase epitaxy of GaN is presented. The governing flow, energy, and species conservation equations are solved in two dimensions to examine the growth characteristics as a function of process variables and reactor geometry. The growth rate varies with GaCl composition but independent of NH3 and H2 flow rates. A change in carrier gas for Ga source from H2 to N2 affects the growth rate and uniformity for a fixed reactor configuration. The model predictions are in general agreement with observed experimental behavior.

  4. Sensitivity analysis of the spectra of the core neutronic source in the calculation of radiation damage in internal of PWR reactor vessel. Internal; Analisis de sensibilidad a los espectros de la fuente neutronica del nucleo en el calculo del dano por irradiacion en los internos de la vasija de un reactor PWR

    Energy Technology Data Exchange (ETDEWEB)

    Cadenas Mendicoa, A. M.; Benito Hernandez, M.; Barrerira Pereira, P.

    2012-07-01

    This study is to analyze the sensitivity to the expected differences in the energy spectra characterizing the neutron source that radiates the vessel internals of a commercial PWR reactor, in order to quantify their influence in the quantities that determine the damage in materials metal.

  5. The changes of the structural, magnetic, and mechanical properties in a reactor pressure vessel steel neutron-irradiated at 70 .deg. C

    CERN Document Server

    Park, D G; Jang, K S; Jung, M M; Kim, G M

    1999-01-01

    The irradiation embrittlement of reactor-pressure-vessel steel has been one of the main safety concerns in nuclear power plants. In the present study, an SA508-3 RPV steel was irradiated by neutrons with various fluences up to 10 sup 1 sup 8 n/cm sup 2 (E>=1MeV) at a temperature of approximately 70 .deg. C. The irradiation responses of the structural, the magnetic, and the mechanical properties of the steel were investigated by means of X-ray diffraction, Moessbauer spectroscopy, magnetic Barkhausen noise, and micro-Vickers hardness measurements. The transitions of all of these parameters occurred above a neutron does of 10 sup 1 sup 6 n/cm sup 2. The results of the X-ray and the Moessbauer experiments revealed that neutron irradiation led to the possibility of partial amorphization in the investigated RPV steel. The changes of the physical and the mechanical properties were discussed in terms of irradiation-induced cascade damage of crystalline materials.

  6. Effect of high-temperature water and hydrogen on the fracture behavior of a low-alloy reactor pressure vessel steel

    Science.gov (United States)

    Roychowdhury, S.; Seifert, H.-P.; Spätig, P.; Que, Z.

    2016-09-01

    Structural integrity of reactor pressure vessels (RPV) is critical for safety and lifetime. Possible degradation of fracture resistance of RPV steel due to exposure to coolant and hydrogen is a concern. In this study tensile and elastic-plastic fracture mechanics (EPFM) tests in air (hydrogen pre-charged) and EFPM tests in hydrogenated/oxygenated high-temperature water (HTW) was done, using a low-alloy RPV steel. 2-5 wppm hydrogen caused embrittlement in air tensile tests at room temperature (25 °C) and at 288 °C, effects being more significant at 25 °C and in simulated weld coarse grain heat affected zone material. Embrittlement at 288 °C is strain rate dependent and is due to localized plastic deformation. Hydrogen pre-charging/HTW exposure did not deteriorate the fracture resistance at 288 °C in base metal, for investigated loading rate range. Clear change in fracture morphology and deformation structures was observed, similar to that after air tests with hydrogen.

  7. Characterisation of the fracture properties in the ductile to brittle transition region of the weld material of a reactor pressure vessel

    Energy Technology Data Exchange (ETDEWEB)

    Scibetta, M. [SCK-CEN, Boeretang 200, 2400 Mol (Belgium); Ferreno, D., E-mail: ferrenod@unican.es [University of Cantabria, ETS Ingenieros de Caminos, Av/Los Castros s/n, 39005 Santander (Spain); Gorrochategui, I. [Centro Tecnologico de Componentes (CTC), Parque Cientifico y Tecnologico de Cantabria, Isabel Torres No 1, 39011 Santander (Spain); Nuclenor, SA, C/Hernan Cortes 26, 39003 Santander (Spain); Lacalle, R. [University of Cantabria, ETS Ingenieros de Caminos, Av/Los Castros s/n, 39005 Santander (Spain); Walle, E. van [SCK-CEN, Boeretang 200, 2400 Mol (Belgium); Martin, J. [Nuclenor, SA, C/Hernan Cortes 26, 39003 Santander (Spain); Gutierrez-Solana, F. [University of Cantabria, ETS Ingenieros de Caminos, Av/Los Castros s/n, 39005 Santander (Spain)

    2011-04-15

    This work presents the results of the fracture characterisation of the weld material of a nuclear vessel, currently in service, in the ductile to brittle transition region. The tests consisted of Charpy impact and tensile tests, performed in the framework of the surveillance programme of the plant. Moreover, in the context of this research, K{sub Jc} fracture toughness tests on pre-cracked Charpy V notch specimens (evaluated according to the Master Curve methodology) together with some mini-tensile tests, were performed; non-irradiated and several irradiated material conditions were characterised. The analysis of the experimental results revealed some inconsistencies concerning the material embrittlement as measured through Charpy and K{sub Jc} fracture tests: in order to obtain an adequate understanding of the results, an extended experimental scope well beyond the regulatory framework was developed, including Charpy tests and K{sub Jc} fracture tests, both performed on reconstituted specimens. Moreover, Charpy specimens irradiated in the high flux BR2 material test reactor were tested with the same purpose. With this extensive experimental programme, a coherent and comprehensive description of the irradiation behaviour of the weld material in the transition region was achieved. Furthermore it revealed better material properties in comparison with the initial expectations based on the information obtained in the framework of the surveillance programme.

  8. Design and fabrication report on instrumented capsule (99M-01K.02H) for korean reactor pressure vessel material made by HANJUNG (Co)

    Energy Technology Data Exchange (ETDEWEB)

    Choo, K. N.; Kang, Y. H.; Kim, B. G.; Cho, M. S.; Son, J. M.; Kim, D. S.; Oh, J. M.; Park, S. J.; Shin, Y. T

    2000-09-01

    The instrumented capsules (99M-01K{center_dot}02H) was designed and fabricated. The purpose of the capsules were to evaluate the nuclear irradiation performance of the Korean nuclear reactor pressure vessel material, SA508 class 3 steel, fabricated by HANJUNG Co for Yonggwang Units 4,5 and Ulchin Unit 4. There are 5 stages having specimens and independent electric heaters in the capsule mainbody. 12 K-type thermocouples and 5 sets of Ni-Ti-Fe and sapphire neutron Fluence Monitors were also inserted in the apsule. Various types of specimens, such as round compact tension, Charpy insert, pre-cracked v-notch (PCVN), tensile, small punch (SP), magnetic Barkhausen effect (MBE), and transmission electron micrograph (TEM) specimens, were inserted in the capsule. The capsule was fabricated at DAEWOO Precision Co. according to KAERI detailed design specifications. This report describes the details of the design, fabrication and inspection of the 99M-01K and 99M-02H capsule. The capsules were irradiated in the IR2 test hole of HANARO at 290{+-}10 deg C up to the fast neutron fluence (E>1.0 MeV) of 3.0x10{sup 19} (n/cm{sup 2})

  9. Characterization of ductile-brittle transition behavior in Mn-Mo-Ni low alloy steels for reactor pressure vessel by small punch test

    Energy Technology Data Exchange (ETDEWEB)

    Kim, M. C.; Lee, B. S.; Park, S. D.; Kim, K. B

    2004-12-01

    Small Punch (SP) tests were performed to evaluate the ductile-brittle transition behavior of Reactor Pressure Vessel (RPV) steels produced by different manufacturing (refining) processes. The results were compared to the standard test results, such as the conventional Charpy impact tests, JIC test and the Master Curve fracture toughness tests in accordance with the ASTM standard E1921. The SP transition temperatures (TSP), which are determined by the middle of the upper SP energy, showed a good correlation with the Charpy index temperatures. And the critical fracture strength, {sigma}{sup *}{sub f(sp)} from small punch test were found to have a linear relationship to the values from the pre-cracked specimens({sigma}{sup *}{sub f(PCVN)}). From the observation of thickness changes according to displacement, Equivalent strain(Eq) could be obtained as a function of ball displacement. It is found that relation between specimen thickness and ball displacement is not dependent on material properties and it is expressed as a function of ball displacement, and then SP equivalent strain(Eq) have close relationship with the fracture toughness(J{sub IC})

  10. Simulation and tests to individual and coupled models of the reactor vessel simulator and the recirculation system for the SUN-RAH; Simulacion y pruebas a modelos individuales y acoplados del simulador de la vasija del reactor y el sistema de recirculacion para el SUN-RAH

    Energy Technology Data Exchange (ETDEWEB)

    Sanchez S, R.A. [UNAM, Laboratorio de Analisis de Ingenieria de Reactores Nucleares, DEPFI, Campus Morelos, en IMTA Jiutepec, Morelos (Mexico)]. e-mail: rsanchez_15@yahoo.com.mx

    2004-07-01

    The present project, is continuation of the project presented in the congress SNM-2003. In this new phase of the project, they were carried out adaptive changes to the modeling and implementation of the module of the full superior of the core of the reactor, they were carried out those modeling of the generation of heat as well as of the energy transfer in the one fuel. These models present the main characteristics of the vessel of the one reactor and of the recirculation system, defined by the main phenomena that they intervene in the physical processes, in the previous version the simulation in real time it required of an extremely quick computer and without executing collateral processes. The tests are presented carried out to the different models belonging to the Simulator of the Reactor Vessel and the Recirculation system for the SUN-RAH (University Simulator of Nucleo electric with Boiling Water Reactor), as well as the results hurtled by this tests. In each section the executions of the tests and the corresponding analyses of results are shown for each pattern. Besides the above mentioned, the advantages presented by the Simulator of the reactor vessel and the recirculation system are pointed. (Author)

  11. NUCLEAR REACTOR

    Science.gov (United States)

    Miller, H.I.; Smith, R.C.

    1958-01-21

    This patent relates to nuclear reactors of the type which use a liquid fuel, such as a solution of uranyl sulfate in ordinary water which acts as the moderator. The reactor is comprised of a spherical vessel having a diameter of about 12 inches substantially surrounded by a reflector of beryllium oxide. Conventionnl control rods and safety rods are operated in slots in the reflector outside the vessel to control the operation of the reactor. An additional means for increasing the safety factor of the reactor by raising the ratio of delayed neutrons to prompt neutrons, is provided and consists of a soluble sulfate salt of beryllium dissolved in the liquid fuel in the proper proportion to obtain the result desired.

  12. CO2 hydrogenation on a metal hydride surface.

    Science.gov (United States)

    Kato, Shunsuke; Borgschulte, Andreas; Ferri, Davide; Bielmann, Michael; Crivello, Jean-Claude; Wiedenmann, Daniel; Parlinska-Wojtan, Magdalena; Rossbach, Peggy; Lu, Ye; Remhof, Arndt; Züttel, Andreas

    2012-04-28

    The catalytic hydrogenation of CO(2) at the surface of a metal hydride and the corresponding surface segregation were investigated. The surface processes on Mg(2)NiH(4) were analyzed by in situ X-ray photoelectron spectroscopy (XPS) combined with thermal desorption spectroscopy (TDS) and mass spectrometry (MS), and time-of-flight secondary ion mass spectrometry (ToF-SIMS). CO(2) hydrogenation on the hydride surface during hydrogen desorption was analyzed by catalytic activity measurement with a flow reactor, a gas chromatograph (GC) and MS. We conclude that for the CO(2) methanation reaction, the dissociation of H(2) molecules at the surface is not the rate controlling step but the dissociative adsorption of CO(2) molecules on the hydride surface.

  13. Research on the Stress Calculation of Reactor Vessel Internal Upper Support Assembly under Accident Case%堆内构件上支承组件在事故工况下的应力计算

    Institute of Scientific and Technical Information of China (English)

    赵文清

    2014-01-01

    Reactor vessel internal upper support assembly adopt shell element combined with beam ele-ment to create model,shell element to create model and solid element to create model,which carry through respective finite element analysis calculation and stress evaluation.The stress calculation effect of reactor vessel internal upper support assembly adopting solid element model created is conservative and high precision by means of modification and simplification about preexistent stress analysis method related to reactor vessel internal,which can satisfy the demand of RCC-M criterion.The calculation method of reactor vessel internal upper support assembly adopting solid element to create model is simplification and practicality by comparing with different creating model mode about stress calculation mentioned above, which may apply to the stress analysis and evaluation of reactor vessel internal assembly related to differ-ent reactor core.%堆内构件上支承组件采用不同的建模方法,分别采用壳单元和梁单元相组合的建模模式、壳单元和壳单元相组合的建模模式、实体单元建模的模式,对堆内构件上支承组件进行了有限元应力计算,比较了不同建模模式下应力计算的各自特点,堆内构件上支承组件实体单元建模模式应力计算结果精确并能满足RC C-M规范应力评定要求,壳单元和梁单元相组合的建模模式、壳单元和壳单元相组合的建模模式应力计算结果保守且应力评定需等效处理其计算结果。堆内构件上支承组件采用整体实体单元全模型建模的计算方法,计算精确且应力评定简单直接,它可应用于其他工况和不同堆芯堆内构件应力计算及其应力评定。

  14. Hydride precipitation kinetics in Zircaloy-4 studied using synchrotron X-ray diffraction

    Science.gov (United States)

    Courty, Olivier F.; Motta, Arthur T.; Piotrowski, Christopher J.; Almer, Jonathan D.

    2015-06-01

    As a result of in-reactor corrosion during operation in nuclear reactors, hydrogen can enter the zirconium fuel cladding and precipitate as brittle hydride particles, which may reduce cladding ductility. Dissolved hydrogen responds to temperature gradients, resulting in transport and precipitation into cold spots so that the distribution of hydrides in the cladding is inhomogeneous. The hydrogen precipitation kinetics plays a strong role in the spatial distribution of the hydrides in the cladding. The precipitation rate is normally described as proportional to the supersaturation of hydrogen in solid solution. The proportionality constant, α2, for hydride precipitation in Zircaloy-4 is measured directly using in situ synchrotron X-Ray diffraction, at different temperatures and with three different initial hydrogen concentrations. The results validate the linear approximation of the phenomenological model and a near constant value of α2 = 4.5 × 10-4 s-1 was determined for the temperature range studied.

  15. Research progress on assessment of reactor vessel integrity under severe accident conditions%严重事故条件下压力容器完整性评价的研究进展

    Institute of Scientific and Technical Information of China (English)

    文青龙; 陈军; 卢冬华; 赵华

    2011-01-01

    堆芯熔融物堆内滞留(In-Vessel Retention,IVR)是以AP1000为代表的第三代轻水反应堆严重事故管理的重要策略之一,也是严重事故条件下保证压力容器完整性(Reactor Vessel Integrity,RVI)的典型方法之一.该文综述了国外在严重事故条件下压力容器完整性试验研究和理论分析的现状,总结了相关的试验装置、试验方法以及基于试验数据拟合得到的经验关联式,评价了严重事故条件下压力容器完整性数值分析的工具和方法,以第三代压水堆热工水力技术为工程背景,探讨了严重事故条件下压力容器完整性热工水力基础研究的方向.%As a representative method of reactor vessel integrity (RVI) under severe accident conditions, In-vessel retention of molten core debris (IVR) is an important severe accident management strategy employed in the API000 generation-3 Pressuried Water Reactor. In this paper, research progress on the test and theoretical analysis based on RVI is reviewed. Test facilities and techniques, as well as the modeling are summarized. In addition, tools for numerical simulation for RVI are evaluated. Finally, based on the applications in thermal hydraulic technology for the generation-3 Pressuried Water Reactor in China, the potential research direction of thermal-hydraulics under RVI conditions are discussed.

  16. Air and metal hydride battery

    Energy Technology Data Exchange (ETDEWEB)

    Lampinen, M.; Noponen, T. [Helsinki Univ. of Technology, Otaniemi (Finland). Lab. of Applied Thermodynamics

    1998-12-31

    The main goal of the air and metal hydride battery project was to enhance the performance and manufacturing technology of both electrodes to such a degree that an air-metal hydride battery could become a commercially and technically competitive power source for electric vehicles. By the end of the project it was possible to demonstrate the very first prototype of the air-metal hydride battery at EV scale, achieving all the required design parameters. (orig.)

  17. Multidimensional simulations of hydrides during fuel rod lifecycle

    Science.gov (United States)

    Stafford, D. S.

    2015-11-01

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

  18. Hydrogen Outgassing from Lithium Hydride

    Energy Technology Data Exchange (ETDEWEB)

    Dinh, L N; Schildbach, M A; Smith, R A; Balazs1, B; McLean II, W

    2006-04-20

    Lithium hydride is a nuclear material with a great affinity for moisture. As a result of exposure to water vapor during machining, transportation, storage and assembly, a corrosion layer (oxide and/or hydroxide) always forms on the surface of lithium hydride resulting in the release of hydrogen gas. Thermodynamically, lithium hydride, lithium oxide and lithium hydroxide are all stable. However, lithium hydroxides formed near the lithium hydride substrate (interface hydroxide) and near the sample/vacuum interface (surface hydroxide) are much less thermally stable than their bulk counterpart. In a dry environment, the interface/surface hydroxides slowly degenerate over many years/decades at room temperature into lithium oxide, releasing water vapor and ultimately hydrogen gas through reaction of the water vapor with the lithium hydride substrate. This outgassing can potentially cause metal hydriding and/or compatibility issues elsewhere in the device. In this chapter, the morphology and the chemistry of the corrosion layer grown on lithium hydride (and in some cases, its isotopic cousin, lithium deuteride) as a result of exposure to moisture are investigated. The hydrogen outgassing processes associated with the formation and subsequent degeneration of this corrosion layer are described. Experimental techniques to measure the hydrogen outgassing kinetics from lithium hydride and methods employing the measured kinetics to predict hydrogen outgassing as a function of time and temperature are presented. Finally, practical procedures to mitigate the problem of hydrogen outgassing from lithium hydride are discussed.

  19. 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

  20. Zirconium hydrides and Fe redistribution in Zr-2.5%Nb alloy under ion irradiation

    Science.gov (United States)

    Idrees, Y.; Yao, Z.; Cui, J.; Shek, G. K.; Daymond, M. R.

    2016-11-01

    Zr-2.5%Nb alloy is used to fabricate the pressure tubes of the CANDU reactor. The pressure tube is the primary pressure boundary for coolant in the CANDU design and is susceptible to delayed hydride cracking, reduction in fracture toughness upon hydride precipitation and potentially hydride blister formation. The morphology and nature of hydrides in Zr-2.5%Nb with 100 wppm hydrogen has been investigated using transmission electron microscopy. The effect of hydrides on heavy ion irradiation induced decomposition of the β phase has been reported. STEM-EDX mapping was employed to investigate the distribution of alloying elements. The results show that hydrides are present in the form of stacks of different sizes, with length scales from nano- to micro-meters. Heavy ion irradiation experiments at 250 °C on as-received and hydrided Zr-2.5%Nb alloy, show interesting effects of hydrogen on the irradiation induced redistribution of Fe. It was found that Fe is widely redistributed from the β phase into the α phase in the as-received material, however, the loss of Fe from the β phase and subsequent precipitation is retarded in the hydrided material. This preliminary work will further the current understanding of microstructural evolution of Zr based alloys in the presence of hydrogen.

  1. Experimental comparison on heat transfer-enhancing component of metal hydride bed

    Energy Technology Data Exchange (ETDEWEB)

    Kang, Hyun-goo, E-mail: hgkang@nfri.re.kr; Chung, Dong-you; Oh, Yun Hee; Chang, Min Ho; Yun, Sei-Hun

    2016-11-01

    Highlights: • Two small ZrCo metal hydride beds were developed. • Copper foam or fin as heat transfer-enhancing component are experimentally compared. • Copper foam bed is more efficient for uniform and rapid heating of metal hydride. • Copper foam bed is more efficient in removal of reaction heat during absorption. - Abstract: Metal hydride bed will be one of the key components for safe handling of tritium in fusion fuel cycle. In case of normal or emergency shutdown of fuel cycle, metal hydride bed installed in storage and delivery system (SDS) of tritium plant will absorb tritium gas in the system as soon as possible. Supply of hydrogen isotope gas to fueling system of fusion reactor will start from the metal hydride beds. Rapid delivery, rapid recovery including rapid heating and cooling are key issues. For better performance of metal hydride bed, various forms of heat transfer enhancing component or design can be applied. This study aims to help the selection of heat transfer enhancing component. Two small ZrCo beds with copper foam and copper fin were developed and experimented with hydrogen gas. Recovery and delivery performance, heating and cooling performance are compared. Experimental results show metal hydride bed with copper foam has improved performance. Uniform heating of metal hydride during desorption and removal of reaction heat during absorption are more efficient with copper foam bed than copper fin bed.

  2. Effect of high-temperature water and hydrogen on the fracture behavior of a low-alloy reactor pressure vessel steel

    Energy Technology Data Exchange (ETDEWEB)

    Roychowdhury, S., E-mail: sroy27@gmail.com [Paul Scherrer Institut, Nuclear Energy and Safety Research Department, Laboratory for Nuclear Materials, 5232 Villigen, PSI (Switzerland); Materials Processing & Corrosion Engineering Division, Mod-Lab, D-Block, Bhabha Atomic Research Centre, Mumbai 400085 (India); Seifert, H.-P.; Spätig, P.; Que, Z. [Paul Scherrer Institut, Nuclear Energy and Safety Research Department, Laboratory for Nuclear Materials, 5232 Villigen, PSI (Switzerland)

    2016-09-15

    Structural integrity of reactor pressure vessels (RPV) is critical for safety and lifetime. Possible degradation of fracture resistance of RPV steel due to exposure to coolant and hydrogen is a concern. In this study tensile and elastic-plastic fracture mechanics (EPFM) tests in air (hydrogen pre-charged) and EFPM tests in hydrogenated/oxygenated high-temperature water (HTW) was done, using a low-alloy RPV steel. 2–5 wppm hydrogen caused embrittlement in air tensile tests at room temperature (25 °C) and at 288 °C, effects being more significant at 25 °C and in simulated weld coarse grain heat affected zone material. Embrittlement at 288 °C is strain rate dependent and is due to localized plastic deformation. Hydrogen pre-charging/HTW exposure did not deteriorate the fracture resistance at 288 °C in base metal, for investigated loading rate range. Clear change in fracture morphology and deformation structures was observed, similar to that after air tests with hydrogen. - Highlights: • Hydrogen content, microstructure of LAS, and strain rate affects tensile properties at 288 °C. • Strength affects hydrogen embrittlement susceptibility to a greater extent than grain size. • Hydrogen in LAS leads to strain localization and restricts cross-slip at 288 °C. • Possible hydrogen pickup due to exposure to 288 °C water alters fracture surface appearance without affecting fracture toughness in bainitic base material. • Simulated weld heat affected zone microstructure shows unstable crack propagation in 288 °C water.

  3. Dual-mode chemical vapor generation for simultaneous determination of hydride-forming and non-hydride-forming elements by atomic fluorescence spectrometry.

    Science.gov (United States)

    Wang, Yu; Xu, Kailai; Jiang, Xiaoming; Hou, Xiandeng; Zheng, Chengbin

    2014-05-21

    A dual-mode chemical vapor generation integrating hydride generation and photochemical vapor generation was developed for simultaneous multi-element analysis of hydride-forming and non-hydride-forming elements by atomic fluorescence spectrometry. Four elements were selected as model elements of hydride-forming (As, Cd) and non-hydride-forming (Ni, Fe) elements to validate this proposed method. Standard or sample solutions were separately pumped to mix with tetrahydroborate, and concentrated formic acid and ammonia, and then directed to a hydride generator and a photochemical reactor to realize simultaneous hydride generation and photochemical vapor generation, respectively. Optimum conditions for dual-mode chemical vapor generation were carefully investigated. Under the optimized conditions, limits of detection of 0.05, 0.008, 0.8 and 0.1 μg L(-1) were obtained for As, Cd, Fe and Ni, respectively. The precisions were 5.0, 5.5, 4.3 and 4.5% (n = 6, RSDs) for 2 μg L(-1) of As, 1 μg L(-1) of Cd, 50 μg L(-1) of Fe and 10 μg L(-1) of Ni, respectively. This method was validated for accuracy with three certified reference water samples and applied to the simultaneous determination of these elements in a tap water sample with spike recoveries in the range of 95-99%.

  4. LMFBR type reactor

    Energy Technology Data Exchange (ETDEWEB)

    Kanbe, Mitsuru

    1997-04-04

    An LMFBR type reactor comprises a plurality of reactor cores in a reactor container. Namely, a plurality of pot containing vessels are disposed in the reactor vessel and a plurality of reactor cores are formed in a state where an integrated-type fuel assembly is each inserted to a pot, and a coolant pipeline is connected to each of the pot containing-vessel to cool the reactor core respectively. When fuels are exchanged, the integrated-type fuel assembly is taken out together with the pot from the reactor vessel in a state where the integrated-type fuel assembly is immersed in the coolants in the pot as it is. Accordingly, coolants are supplied to each of the pot containing-vessel connected with the coolant pipeline and circulate while cooling the integrated-type fuel assembly for every pot. Then, when the fuels are exchanged, the integrated type fuel assembly is taken out to the outside of the reactor together with the pot by taking up the pot from the pot-containing vessel. Then, neutron economy is improved to thereby improve reactor power and the breeding ratio. (N.H.)

  5. Description of a disposition line on the stress corrosion cracking behaviour of ferritic reactor pressure vessel steels under BWR-conditions; Beschreibung einer einhuellenden Risswachstumskurve zum Spannungsrisskorrosionsverhalten von ferritischen Reaktordruckbehaelter (RDB)-Staehlen unter Siedewasserreaktor (SWR)-Bedingungen

    Energy Technology Data Exchange (ETDEWEB)

    Bruemmer, G. [HEW, Hamburg (Germany); Hoffmann, H. [VGB-GS, Essen (Germany); Ilg, U. [EnBW Kraftwerke AG, Philippsburg (Germany); Wachter, O. [E.ON-Kernkraft, Hannover (Germany); Widera, M. [RWE Power, Essen (Germany); Roth, A. [Framatome ANP GmbH, Erlangen (Germany)

    2002-07-01

    The inner surface of the reactor pressure vessel of BWR reactors is lined with a welded, corrosion-resistant steel liner. In an assumed case of liner rupture down to the low-alloy ferritic base material, an integrity assessment of the pressure vesssel in consideration of the effects of reactor coolant is of utmost importance, and research in this field has been going on for more than ten years now. An analysis of the available data shows that it is now possible to describe a disposition line on the stress corrosion cracking behaviour of ferritic reactor pressure vessel steels in BWR conditions. Crack growth rates of a stress intensity factor corresponding to a T/4 wall defect (i.e. 25 percent of the wall thickness) are technically not relevant. This scientific finding is supported by measurements of about 450 reactor operation years of all German LWR reactor plants, none of which showed crack initiation in the reactor pressure vessel. [German] Die mediumberuehrte Innenoberflaeche des Reaktordruckbehaelters (RDB) von Siedewasserreaktoren (SWR) ist mit einer korrosionsbestaendigen austenitischen Schweissplattierung versehen. Fuer den unterstellten Fall einer bis auf den niedriglegierten, ferritischen Grundwerkstoff durchgerissenen Pattierung ist fuer die Beurteilung der Integritaet des RDB unter Beruecksichtigung der Einwirkung des Reaktorkuehlmittels die Klaerung der Frage eines korrosionsgestuetzten Risswachstums von grosser Bedeutung. Dieses Thema ist daher bereits seit mehr als 10 Jahren Gegenstand umfangreicher Forschungsaktivitaeten. Ende der 80er- und Anfang der 90er-Jahre wurden fuer ferritische RDB-Staehle von SWR-Anlagen Risswachstumsgeschwindigkeiten veroeffentlicht, die binnen weniger als einem Jahr zum Durchriss der drucktragenden Wand eines RDB gefuehrt haetten. Daraufhin wurden internationale Forschungsaktivitaeten zur Ermittlung zuverlaessiger und reproduzierbarer Risswachstumsdaten initiiert, deren Ergebnisse zusammenfassend dargestellt werden. Die

  6. Estimate of radiation-induced steel embrittlement in the BWR core shroud and vessel wall from reactor-grade MOX/UOX fuel for the nuclear power plant at Laguna Verde, Veracruz, Mexico

    Science.gov (United States)

    Vickers, Lisa Rene

    The government of Mexico has expressed interest to utilize the Laguna Verde boiling water reactor (BWR) nuclear power plant for the disposition of reprocessed spent uranium oxide (UOX) fuel in the form of reactor-grade mixed-oxide (MOX) fuel. MOX fuel would replace spent UOX fuel as a fraction in the core from 18--30% depending on the fuel loading cycle. MOX fuel is expected to increase the neutron fluence, flux, fuel centerline temperature, reactor core pressure, and yield higher energy neutrons. There is concern that a core with a fraction of MOX fuel (i.e., increased 239Pu wt%) would increase the radiation-induced steel embrittlement within the core shroud and vessel wall as compared to only conventional, enriched UOX fuel in the core. The evaluation of radiation-induced steel embrittlement within the core shroud and vessel wall is a concern because of the potentially adverse affect to plant and public safety, environment, and operating life of the reactor. This dissertation provides computational results of the neutron fluence, flux, energy spectrum, and radiation damage displacements per atom per second (dpa-s-1) in steel within the core shroud and vessel wall of the Laguna Verde Unit 1 BWR. The results were computed using the nuclear data processing code NJOY99 and the continuous energy Monte Carlo Neutral Particle transport code MCNP4B. The MCNP4B model of the reactor core was for maximum core loading fractions of ⅓ MOX and ⅔ UOX reactor-grade fuel in an equilibrium core. The primary conclusion of this dissertation was that the addition of the maximum fraction of ⅓ MOX fuel to the LV1 BWR core did significantly accelerate the radiation-induced steel embrittlement such that without mitigation of steel embrittlement by periodic thermal annealing or reduction in operating parameters such as, neutron fluence, core temperature and pressure, it posed a potentially adverse affect to the plant and public safety, environment, and operating life of the reactor.

  7. Pressure vessel design manual

    CERN Document Server

    Moss, Dennis R

    2013-01-01

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

  8. Performance study of a hydrogen powered metal hydride actuator

    Science.gov (United States)

    Mainul Hossain Bhuiya, Md; Kim, Kwang J.

    2016-04-01

    A thermally driven hydrogen powered actuator integrating metal hydride hydrogen storage reactor, which is compact, noiseless, and able to generate smooth actuation, is presented in this article. To test the plausibility of a thermally driven actuator, a conventional piston type actuator was integrated with LaNi5 based hydrogen storage system. Copper encapsulation followed by compaction of particles into pellets, were adopted to improve overall thermal conductivity of the reactor. The operation of the actuator was thoroughly investigated for an array of operating temperature ranges. Temperature swing of the hydride reactor triggering smooth and noiseless actuation over several operating temperature ranges were monitored for quantification of actuator efficiency. Overall, the actuator generated smooth and consistent strokes during repeated cycles of operation. The efficiency of the actuator was found to be as high as 13.36% for operating a temperature range of 20 °C-50 °C. Stress-strain characteristics, actuation hysteresis etc were studied experimentally. Comparison of stress-strain characteristics of the proposed actuator with traditional actuators, artificial muscles and so on was made. The study suggests that design modification and use of high pressure hydride may enhance the performance and broaden the application horizon of the proposed actuator in future.

  9. Influence of hydrides orientation on strain, damage and failure of hydrided zircaloy-4; Influence de l'orientation des hydrures sur les modes de deformation, d'endommagement et de rupture du zircaloy-4 hydrure

    Energy Technology Data Exchange (ETDEWEB)

    Racine, A

    2005-09-15

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

  10. Study and Realization of Auto Ultrasonic-testing Technology for Reactor Pressure Vessel Studs%核电厂主螺栓超声自动检测技术研究与实现

    Institute of Scientific and Technical Information of China (English)

    张宝军; 张国丰; 严智; 刘呈则

    2013-01-01

    Reactor pressure vessel studs applied in the environment of high temperature , high pressure , and high radiation ,are easy to be suffered from fatigue damage .A set of auto ultrasonic -testing system of the reactor pressure vessel studs is designed .The main composition ,function ,and detection technology of the inspection system were introduced .It′s confirmed by experiment that the system and technology meets the requirement of ASME code volume Ⅺon the inspection of reactor pressure vessel studs of pressurized water reactor nuclear power station during PSI ( pre -service inspection ) and ISI ( in-service inspec-tion) .By developing the system and studying on the technology ,the automation inspection of reactor pres-sure vessel studs is realized .And at the same time ,the detection efficiency and quality is improved .This system also can be applied to other different reactor′s studs inspection and civilian pressure vessel′s large bolts inspection .%  核电厂反应堆压力容器主螺栓长期运行在高温、高压、高放射性环境下,易于形成疲劳损伤。针对核电厂反应堆压力容器主螺栓的检测要求,研究并形成了一套主螺栓超声检测系统及检测技术。介绍了该检测系统的主要组成、功能及相应的检测技术,通过试验方法验证该系统及技术满足ASME规范第Ⅺ卷关于压水堆核电站役前和在役检查反应堆压力容器主螺栓体积性检查的规定要求。该系统及技术的研究,实现了核电厂反应堆压力容器主螺栓检测过程的自动化,提高了检测效率,可适用多种堆型的核电厂主螺栓检查以及其他民用承压大型螺栓的检查,具有广泛的适用性。

  11. Ex-vessel neutron dosimetry analysis for westinghouse 4-loop XL pressurized water reactor plant using the RadTrack{sup TM} Code System with the 3D parallel discrete ordinates code RAPTOR-M3G

    Energy Technology Data Exchange (ETDEWEB)

    Chen, J.; Alpan, F. A.; Fischer, G.A.; Fero, A.H. [Westinghouse Electric Company, Nuclear Services, Radiation Engineering and Analysis, 1000 Westinghouse Dr., Cranberry Township, PA 16066-5228 (United States)

    2011-07-01

    Traditional two-dimensional (2D)/one-dimensional (1D) SYNTHESIS methodology has been widely used to calculate fast neutron (>1.0 MeV) fluence exposure to reactor pressure vessel in the belt-line region. However, it is expected that this methodology cannot provide accurate fast neutron fluence calculation at elevations far above or below the active core region. A three-dimensional (3D) parallel discrete ordinates calculation for ex-vessel neutron dosimetry on a Westinghouse 4-Loop XL Pressurized Water Reactor has been done. It shows good agreement between the calculated results and measured results. Furthermore, the results show very different fast neutron flux values at some of the former plate locations and elevations above and below an active core than those calculated by a 2D/1D SYNTHESIS method. This indicates that for certain irregular reactor internal structures, where the fast neutron flux has a very strong local effect, it is required to use a 3D transport method to calculate accurate fast neutron exposure. (authors)

  12. Chemical Hydride Slurry for Hydrogen Production and Storage

    Energy Technology Data Exchange (ETDEWEB)

    McClaine, Andrew W

    2008-09-30

    hydrogen. Further evaluation and development of this concept will be performed as follow-on work under another project. However, since the cost of reducing magnesium from magnesium oxide makes up 85% of the cost of the slurry, if hydrogen can be stored many times in the slurry, then the cost of storing hydrogen can be spread over many units of hydrogen and can be significantly reduced from the costs of a chemical hydride system. This may be the most important finding of this project. If the slurry is used to carry a rechargeable hydride, the slurry can be stored in a conventional liquid fuel tank and delivered to a release system as hydrogen is needed. The release system will contain only the hydride needed to produce the hydrogen desired. This is in contrast to conventional designs proposed for other rechargeable hydride systems that store all the hydride in a large and heavy pressure and heat transfer vessel.

  13. JAERI's contribution to the IAEA coordinated research programme on assuring structural integrity of reactor pressure vessels' (CRP-IV). Final report (Contract research)

    Energy Technology Data Exchange (ETDEWEB)

    Onizawa, Kunio; Suzuki, Masahide [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    2000-05-01

    According to the Research Agreement No. 9736 between the IAEA and the JAERI, we commenced the test program for the IAEA Coordinated Research Program (CRP) on 'Assuring Structural Integrity of Reactor Pressure Vessels' at JAERI in September 1997. For the program, we received one block of the IAEA reference material JRQ from the IAEA CRP coordinator in June 1997. The test program has been conducted using the JRQ block and additional materials (Steels A and B) from our own program having a similar object with the CRP. The CRP consists of two parts; a mandatory part and an optional part. For the mandatory part of the JAERI program, instrumented Charpy impact tests and fracture toughness tests using precracked Charpy-v (PCCv) specimens were performed. As the optional part, neutron irradiation to specimens of JRQ was conducted at JMTR by using two capsules. In this report, the results of the mandatory part and irradiated Charpy and PCCv specimens of JRQ from capsule No.1 as well as those of Steel A and Steel B were described. The following conclusions were drawn; (1) the data form Charpy impact and fracture toughness tests of JRQ agreed well with the data in the CRP-3, (2) the scatter of fracture toughness of JRQ is relatively large, i. e., the Weibull slope 'b' is less than 3, (3) the reference temperature T{sub 0} from PCCv is in good agreement with T{sub 0} from 1T-Compact Tension (CT) when the tests are performed at the recommended temperature or the data has no invalid data, (4) the reference temperature T{sub 0} after neutron irradiation can be determined with six to eight specimens at the recommended temperature and (5) the shift of the reference temperature T{sub 0} is almost equivalent to the shift of Charpy transition temperature, but affected by the treatment of the highest data and testing temperature. Further studies on the fracture toughness evaluation are necessary concerning the treatment of outlier, temperature dependence after

  14. Evaluation of the performance of ultrasonic and eddy current testing of austenitic claddings of reactor pressure vessels; Bewertung der Aussagefaehigkeit von Ultraschall- und Wirbelstrompruefung austenitischer Plattierungen von Reaktordruckbehaeltern

    Energy Technology Data Exchange (ETDEWEB)

    Waidele, H.; Maier, H.J.; Just, T.; Seidenkranz, T.; Seydel, O.; Weiss, R.

    2000-12-01

    In the scope of this project, non-destructive testing methods were carried out on specimens with defects intentionally manufactured in the region of the cladding. The aim of these trials is an evaluation of the performance of ultrasonic and eddy current examinations of austenitic claddings of reactor pressure vessels. A review of the non-destructive testing of claddings showed that the majority of the investigations have been carried out on specimens with artificial defects (notches, holes). Therefore, for the realisation of this project MPA Stuttgart produced specimens with natural defects in the cladding. In detail these are specimens with intergranular stress-corrosion cracking, hot cracks and welding defects in the cladding as well as specimens with underclad cracks. The thickness of the specimens is about 150 mm (BWR-RPV), so that in addition to the testing from the ID (PWR, ultrasonic, eddy current) also the testing from the OD (BWR, ultrasonic) could be examined. The measurements show that most of the cladding defects can be detected with the standard ultrasonic test methods, however, in some cases generate only low echo amplitudes. Favourable results were obtained from the ID testing by means of a phased array probe, in particular in connection with the eddy current technique. Investigations on specimens containing defects not known to the inspection teams (blind tests), which will allow a further evaluation of the performance of non-destructive testing methods under realistic conditions, will be carried out in Phase II of the project. (orig.) [German] Im Rahmen dieses Vorhabens wurden an Testkoerpern mit im Plattierungsbereich gezielt eingebrachten Fehlerbildungen zerstoerungsfreie Pruefungen durchgefuehrt. Ziel der Untersuchungen ist eine Bewertung der Aussagefaehigkeit von Ultraschall- und Wirbelstrompruefungen an austenitischen Plattierungen von Reaktordruckbehaeltern. Bei einer Bestandsaufnahme zur zerstoerungsfreien Pruefung von Plattierungen zeigte

  15. Vessel Operating Units (Vessels)

    Data.gov (United States)

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

  16. A deformation and thermodynamic model for hydride precipitation kinetics in spent fuel cladding

    Energy Technology Data Exchange (ETDEWEB)

    Stout, R.B.

    1989-10-01

    Hydrogen is contained in the Zircaloy cladding of spent fuel rods from nuclear reactors. All the spent fuel rods placed in a nuclear waste repository will have a temperature history that decreases toward ambient; and as a result, most all of the hydrogen in the Zircaloy will eventually precipitate as zirconium hydride platelets. A model for the density of hydride platelets is a necessary sub-part for predicting Zircaloy cladding failure rate in a nuclear waste repository. A model is developed to describe statistically the hydride platelet density, and the density function includes the orientation as a physical attribute. The model applies concepts from statistical mechanics to derive probable deformation and thermodynamic functionals for cladding material response that depend explicitly on the hydride platelet density function. From this model, hydride precipitation kinetics depend on a thermodynamic potential for hydride density change and on the inner product of a stress tensor and a tensor measure for the incremental volume change due to hydride platelets. The development of a failure response model for Zircaloy cladding exposed to the expected conditions in a nuclear waste repository is supported by the US DOE Yucca Mountain Project. 19 refs., 3 figs.

  17. Metal Hydride Compression

    Energy Technology Data Exchange (ETDEWEB)

    Johnson, Terry A. [Sandia National Lab. (SNL-CA), Livermore, CA (United States); Bowman, Robert [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Smith, Barton [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Anovitz, Lawrence [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Jensen, Craig [Hawaii Hydrogen Carriers LLC, Honolulu, HI (United States)

    2017-07-01

    Conventional hydrogen compressors often contribute over half of the cost of hydrogen stations, have poor reliability, and have insufficient flow rates for a mature FCEV market. Fatigue associated with their moving parts including cracking of diaphragms and failure of seal leads to failure in conventional compressors, which is exacerbated by the repeated starts and stops expected at fueling stations. Furthermore, the conventional lubrication of these compressors with oil is generally unacceptable at fueling stations due to potential fuel contamination. Metal hydride (MH) technology offers a very good alternative to both conventional (mechanical) and newly developed (electrochemical, ionic liquid pistons) methods of hydrogen compression. Advantages of MH compression include simplicity in design and operation, absence of moving parts, compactness, safety and reliability, and the possibility to utilize waste industrial heat to power the compressor. Beyond conventional H2 supplies of pipelines or tanker trucks, another attractive scenario is the on-site generating, pressuring and delivering pure H2 at pressure (≥ 875 bar) for refueling vehicles at electrolysis, wind, or solar generating production facilities in distributed locations that are too remote or widely distributed for cost effective bulk transport. MH hydrogen compression utilizes a reversible heat-driven interaction of a hydride-forming metal alloy with hydrogen gas to form the MH phase and is a promising process for hydrogen energy applications [1,2]. To deliver hydrogen continuously, each stage of the compressor must consist of multiple MH beds with synchronized hydrogenation & dehydrogenation cycles. Multistage pressurization allows achievement of greater compression ratios using reduced temperature swings compared to single stage compressors. The objectives of this project are to investigate and demonstrate on a laboratory scale a two-stage MH hydrogen (H2) gas compressor with a

  18. 核能系统压力容器辐照脆化机制及其影响因素%IRRADIATION EMBRITTLEMENT MECHANISMS AND RELEVANT INFLUENCE FACTORS OF NUCLEAR REACTOR PRESSURE VESSEL STEELS

    Institute of Scientific and Technical Information of China (English)

    李正操; 陈良

    2014-01-01

    Nuclear reactor pressure vessel is the irreplaceable component of the nuclear power plant and its integrity is one of the key issues of any nuclear power plant for long term operations.Various nanofeatures,including solute clusters,matrix damage and grain boundary segregation formed in reactor pressure vessel steels in the face of neutron irradiation.These ultrafine microstructural features lead to an increase in the ductile brittle transition temperature as is the measure used to describe the irradiation embrittlement.The balance of features depends on the composition of the reactor pressure vessel steels and the irradiation conditions.This paper reviews the current phenomenological knowledge and understanding of the basic mechanisms and relevant influence factors for irradiation embrittlement of nuclear reactor pressure vessel steels.To be specific,the formation and evolution processes of the embrittling features are presented.Also,the influences of material variables,such as copper,nickel and manganese contents on irradiation embrittlement and those of irradiation variables,such as neutron flux and post irradiation annealing are summarized.In addition,fundamental research issues that remain to be addressed are briefly pointed out.%核反应堆压力容器作为核电站不可更换的关键性设备,其设备完整性对核电站的安全运行起着至关重要的作用.在辐照条件下,反应堆压力容器钢中会形成一系列微结构缺陷,包括溶质沉淀、基体损伤和脆性元素的晶界偏聚等,导致材料的韧脆性转变温度升高,产生辐照脆化效应.而压力容器钢的成分和辐照条件决定了各种微结构对辐照脆化的贡献大小.本文主要针对核能系统压力容器辐照脆化机制及其影响因素进行了综述,总结讨论了这些微结构的形成机制及溶质元素、辐照通量和辐照后退火对这些微结构和材料机械性能的影响,并指出了存在的问题和未来的研究方向.

  19. Production of propylene from 1-butene on highly active "bi-functional single active site" catalyst: Tungsten carbene-hydride supported on alumina

    KAUST Repository

    Mazoyer, Etienne

    2011-12-02

    1-Butene is transformed in a continuous flow reactor over tungsten hydrides precursor W-H/Al2O3, 1, giving a promising yield into propylene at 150 °C and different pressures. Tungsten carbene-hydride single active site operates as a "bi-functional catalyst" through 1-butene isomerization on W-hydride and 1-butene/2-butenes cross-metathesis on W-carbene. This active moiety is generated in situ at the initiation steps by insertion of 1-butene on tungsten hydrides precursor W-H/Al2O3, 1 followed by α-H and β-H abstraction. © 2011 American Chemical Society.

  20. 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

  1. Erbium hydride decomposition kinetics.

    Energy Technology Data Exchange (ETDEWEB)

    Ferrizz, Robert Matthew

    2006-11-01

    Thermal desorption spectroscopy (TDS) is used to study the decomposition kinetics of erbium hydride thin films. The TDS results presented in this report are analyzed quantitatively using Redhead's method to yield kinetic parameters (E{sub A} {approx} 54.2 kcal/mol), which are then utilized to predict hydrogen outgassing in vacuum for a variety of thermal treatments. Interestingly, it was found that the activation energy for desorption can vary by more than 7 kcal/mol (0.30 eV) for seemingly similar samples. In addition, small amounts of less-stable hydrogen were observed for all erbium dihydride films. A detailed explanation of several approaches for analyzing thermal desorption spectra to obtain kinetic information is included as an appendix.

  2. Hydride development for hydrogen storage

    Energy Technology Data Exchange (ETDEWEB)

    Thomas, G.J.; Guthrie, S.E.; Bauer, W.; Yang, N.Y.C. [Sandia National Lab., Livermore, CA (United States); Sandrock, G. [SunaTech, Inc., Ringwood, NJ (United States)

    1996-10-01

    The purpose of this project is to develop and demonstrate improved hydride materials for hydrogen storage. The work currently is organized into four tasks: hydride development, bed fabrication, materials support for engineering systems, and IEA Annex 12 activities. At the present time, hydride development is focused on Mg alloys. These materials generally have higher weight densities for storing hydrogen than rare earth or transition metal alloys, but suffer from high operating temperatures, slow kinetic behavior and material stability. The authors approach is to study bulk alloy additions which increase equilibrium overpressure, in combination with stable surface alloy modification and particle size control to improve kinetic properties. This work attempts to build on the considerable previous research in this area, but examines specific alloy systems in greater detail, with attention to known phase properties and structures. The authors have found that specific phases can be produced which have significantly improved hydride properties compared to previous studies.

  3. Metal hydride air conditioner

    Institute of Scientific and Technical Information of China (English)

    YANG; Ke; DU; Ping; LU; Man-qi

    2005-01-01

    The relationship among the hydrogen storage properties, cycling characteristics and thermal parameters of the metal hydride air conditioning systems was investigated. Based on a new alloy selection model, three pairs of hydrogen storage alloys, LaNi4.4 Mn0.26 Al0.34 / La0.6 Nd0.4 Ni4.8 Mn0.2 Cu0. 1, LaNi4.61Mn0. 26 Al0.13/La0.6 Nd0.4 Ni4.8 Mn0.2 Cu0. 1 and LaNi4.61 Mn0.26 Al0.13/La0.6 Y0.4 Ni4.8 Mn0. 2, were selected as the working materials for the metal hydride air conditioning system. Studies on the factors affecting the COP of the system showed that higher COP and available hydrogen content need the proper operating temperature and cycling time,large hydrogen storage capacity, flat plateau and small hysterisis of hydrogen alloys, proper original input hydrogen content and mass ratio of the pair of alloys. It also needs small conditioning system was established by using LaNi4.61 Mn0.26 Al0. 13/La0.6 Y0.4 Ni4.8 Mn0.2 alloys as the working materials, which showed that under the operating temperature of 180℃/40℃, a low temperature of 13℃ was reached, with COP =0.38 and Wnet =0.09 kW/kg.

  4. The effect of stress state on zirconium hydride reorientation

    Science.gov (United States)

    Cinbiz, Mahmut Nedim

    Prior to storage in a dry-cask facility, spent nuclear fuel must undergo a vacuum drying cycle during which the spent fuel rods are heated up to elevated temperatures of ≤ 400°C to remove moisture the canisters within the cask. As temperature increases during heating, some of the hydride particles within the cladding dissolve while the internal gas pressure in fuel rods increases generating multi-axial hoop and axial stresses in the closed-end thin-walled cladding tubes. As cool-down starts, the hydrogen in solid solution precipitates as hydride platelets, and if the multiaxial stresses are sufficiently large, the precipitating hydrides reorient from their initial circumferential orientation to radial orientation. Radial hydrides can severely embrittle the spent nuclear fuel cladding at low temperature in response to hoop stress loading. Because the cladding can experience a range of stress states during the thermo-mechanical treatment induced during vacuum drying, this study has investigated the effect of stress state on the process of hydride reorientation during controlled thermo-mechanical treatments utilizing the combination of in situ X-ray diffraction and novel mechanical testing analyzed by the combination of metallography and finite element analysis. The study used cold worked and stress relieved Zircaloy-4 sheet containing approx. 180 wt. ppm hydrogen as its material basis. The failure behavior of this material containing radial hydrides was also studied over a range of temperatures. Finally, samples from reactor-irradiated cladding tubes were examined by X-ray diffraction using synchrotron radiation. To reveal the stress state effect on hydride reorientation, the critical threshold stress to reorient hydrides was determined by designing novel mechanical test samples which produce a range of stress states from uniaxial to "near-equibiaxial" tension when a load is applied. The threshold stress was determined after thermo-mechanical treatments by

  5. Geoneutrino and Hydridic Earth model

    CERN Document Server

    Bezrukov, Leonid

    2013-01-01

    Uranium, Thorium and Potassium-40 abundances in the Earth were calculated in the frame of Hydridic Earth model. Terrestrial heat producton from U, Th and K40 decays was calculated also. We must admit the existance of Earth expansion process to understand the obtained large value of terrestrial heat producton. The geoneutrino detector with volume more than 5 kT (LENA type) must be constructed to definitely separate between Bulk Silicat Earth model and Hydridic Earth model.

  6. The vessel fluence; Fluence cuve

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2001-07-01

    This book presents the proceedings of the technical meeting on the reactors vessels fluence. They are grouped in eight sessions: the industrial context and the stakes of the vessels control; the organization and the methodology for the fluence computation; the concerned physical properties; the reference computation methods; the fluence monitoring in an industrial context; vessels monitoring under irradiation; others methods in the world; the research and development programs. (A.L.B.)

  7. Zirconium hydride precipitation kinetics in Zircaloy-4 observed with synchrotron X-ray diffraction

    Energy Technology Data Exchange (ETDEWEB)

    Blackmur, M.S., E-mail: matthew.blackmur@postgrad.manchester.ac.uk [Materials Performance Centre, School of Materials, The University of Manchester, Manchester M1 7HS (United Kingdom); Robson, J.D.; Preuss, M. [Materials Performance Centre, School of Materials, The University of Manchester, Manchester M1 7HS (United Kingdom); Zanellato, O. [PIMM, Ensam – Cnam – CNRS, 151 Boulevard de l’Hôpital, 75013 Paris (France); Cernik, R.J. [Materials Performance Centre, School of Materials, The University of Manchester, Manchester M1 7HS (United Kingdom); Shi, S.-Q. [Department of Mechanical Engineering, The Hong Kong Polytechnic University, Hung Hom, Kowloon, Hong Kong (China); Ribeiro, F. [Institut de Radioprotection et de Sûreté Nucléaire, CEN Cadarache, 13115 St. Paul Les Durance (France); Andrieux, J. [Beamline ID15, European Synchrotron Radiation Facility, Grenoble (France)

    2015-09-15

    High-energy synchrotron X-ray diffraction was used to investigate the isothermal precipitation of δ-hydride platelets in Zircaloy-4 at a range of temperatures relevant to reactor conditions, during both normal operation and thermal transients. From an examination of the rate kinetics of the precipitation process, precipitation slows with increasing temperature above 200 °C, due to a reduction in the thermodynamic driving force. A model for nucleation rate as a function of temperature was developed, to interpret the precipitation rates seen experimentally. While the strain energy associated with the misfit between hydrides and the matrix makes a significant contribution to the energy barrier for nucleation, a larger contribution arises from the interfacial energy. Diffusion distance calculations show that hydrogen is highly mobile in the considered thermal range and on the scale of inter-hydride spacing and it is not expected to be significantly rate limiting on the precipitation process that takes place under reactor operating conditions.

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

    Energy Technology Data Exchange (ETDEWEB)

    Van Blarigan, Peter; Shugard, Andrew D.; Walters, R. Tom (Savannah River National Labs, Aiken, SC)

    2012-01-01

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

  9. Design concept of conducting shell and in-vessel components suitable for plasma vertical stability and remote maintenance scheme in DEMO reactor

    Energy Technology Data Exchange (ETDEWEB)

    Utoh, Hiroyasu, E-mail: uto.hiroyasu@jaea.go.jp [Japan Atomic Energy Agency, Obuchi, Rokkasho-mura, Aomori-ken 039-3212 (Japan); International Fusion Energy Research Centre, 2-166, Obuchi, Rokkasho, Aomori 039-3212 (Japan); Takase, Haruhiko [Japan Atomic Energy Agency, Obuchi, Rokkasho-mura, Aomori-ken 039-3212 (Japan); International Fusion Energy Research Centre, 2-166, Obuchi, Rokkasho, Aomori 039-3212 (Japan); Sakamoto, Yoshiteru; Tobita, Kenji [Japan Atomic Energy Agency, Obuchi, Rokkasho-mura, Aomori-ken 039-3212 (Japan); Mori, Kazuo; Kudo, Tatsuya [Japan Atomic Energy Agency, Obuchi, Rokkasho-mura, Aomori-ken 039-3212 (Japan); International Fusion Energy Research Centre, 2-166, Obuchi, Rokkasho, Aomori 039-3212 (Japan); Someya, Youji; Asakura, Nobuyuki; Hoshino, Kazuo; Nakamura, Makoto; Tokunaga, Shinsuke [Japan Atomic Energy Agency, Obuchi, Rokkasho-mura, Aomori-ken 039-3212 (Japan)

    2016-02-15

    Highlights: • Conceptual design of in-vessel component including conducting shell has been investigated. • The conducting shell design for plasma vertical stability was clarified from the plasma vertical stability analysis. • The calculation results showed that the double-loop shell has the most effect on plasma vertical stability. - Abstract: In order to realize a feasible DEMO, we designed an in-vessel component including the conducting shell. The project is affiliated with the broader approach DEMO design activities and is conceptualized from a plasma vertical stability and engineering viewpoint. The dependence of the plasma vertical stability on the conducing shell parameters and the electromagnetic force at plasma disruption were investigated in numerical simulations (programmed in the 3D eddy current analysis code and a plasma position control code). The simulations assumed the actual shape and position of the vacuum vessel and in-vessel components. The plasma vertical stability was most effectively maintained by the double-loop shell.

  10. Evaluation on the Feasibility of Using Ultrasonic Testing of Reactor Pressure Vessel Welds for Assessing Flaw Density/Distribution per 10 CFR 50.61a, Alternate Fracture Toughness Requirements for Protection Against Pressurized Thermal Shock

    Energy Technology Data Exchange (ETDEWEB)

    Sullivan, Edmund J.; Anderson, Michael T.

    2014-06-10

    This technical letter report provides the status of an assessment undertaken by PNNL at the request of the NRC to verify the capability of periodic ASME-required volumetric examinations of reactor vessels to characterize the density and distribution of flaws of interest for applying §50.61a on a plant-by-plant basis. The PTS rule, described in the Code of Federal Regulations, Title 10, Section 50.61 (§50.61), "Fracture Toughness Requirements for Protection against Pressurized Thermal Shock Events," establishes screening criteria to ensure that the potential for a reactor vessel to fail due to a PTS event is deemed to be acceptably low. Recently, the NRC completed a research program that concluded that the risk of through-wall cracking due to a PTS event is much lower than previously estimated. The NRC subsequently developed and promulgated an alternate PTS rule, §50.61a, that can be implemented by PWR licensees. The §50.61a rule differs from §50.61 in that it requires licensees who choose to follow this alternate method to analyze the results from periodic volumetric examinations required by the ASME Code, Section XI, Rules for Inservice Inspection (ISI) of Nuclear Power Plants.

  11. Stress analysis in a non axisymmetric loaded reactor pressure vessel; Verificacao de tensoes em um vaso de pressao nuclear com carregamentos nao-axissimetricos

    Energy Technology Data Exchange (ETDEWEB)

    Albuquerque, Levi Barcelos; Assis, Gracia Menezes V. de [Coordenadoria para Projetos Especiais (COPESP), Sao Paulo, SP (Brazil); Miranda, Carlos Alexandre J.; Cruz, Julio Ricardo B.; Mattar Neto, Miguel [Instituto de Pesquisas Energeticas e Nucleares (IPEN), Sao Paulo, SP (Brazil)

    1995-12-31

    In this work we intend to present the stress analysis of a PWR vessel under postulated concentrated loads. The vessel was modeled with Axisymmetric solid 4 nodes harmonic finite elements with the use of the ANSYS program, version 5.0. The bolts connecting the vessel flanges were modeled with beam elements. Some considerations were made to model the contact between the flanges. The perforated part of the vessel tori spherical head was modeled (with reduced properties due to its holes) to introduce its stiffness and loads but was not within the scope of this work. The loading consists of some usual ones, as pressure, dead weight, bolts preload, seismic load and some postulated ones as concentrated loads, over the vessel, modeled by Fourier Series. The results in the axisymmetric model are taken in terms of linearized stresses, obtained in some circumferential positions and for each position, in some sections along the vessel. Using the ASME Code (Section III, Division 1, Sub-section NB) the stresses are within the allowable limits. In order to draw some conclusions about stress linearization, the membrane plus bending stresses (Pl + Pb) are obtained and compared in some sections, using three different methods. (author) 4 refs., 15 figs., 7 tabs.

  12. Experimental investigation of heat transfer during severe accident of a Pressurized Heavy Water Reactor with simulated decay heat generation in molten pool inside calandria vessel

    Energy Technology Data Exchange (ETDEWEB)

    Prasad, Sumit Vishnu, E-mail: svprasad@barc.gov.in; Nayak, Arun Kumar, E-mail: arunths@barc.gov.in

    2016-07-15

    Highlights: • Scaled test facility simulating the calandria vessel and calandria vault water of PHWR with simulated decay heat was built. • Experiments conducted with simulant material at about 1200 °C. • Experimental result shows that melt coolability and growth rate of crust thickness are affected by presence of decay heat. • No gap was observed between the crust and vessel on opening. • Result shows that vessel integrity is intact with presence of water inside water tank in both cases. - Abstract: The present study focuses on experimental investigation in a scaled facility of an Indian PHWR to investigate the coolability of molten corium with simulated decay heat in the simulated calandria vessel. Molten borosilicate glass was used as the simulant due to its comparable heat transfer characteristics similar to prototypic material. About 60 kg of the molten material was poured into the test section at about 1200 °C. Decay heat in the melt pool was simulated using four high watt heaters cartridges, each having 9.2 kW. The temperature distributions inside the molten pool, across the vessel wall thickness and vault water were measured. Experimental results obtained are compared with the results obtained previously for no decay heat case. The results indicated that presence of decay heat seriously affects the coolability behaviour and formation of crust in the melt pool. The location and magnitude of maximum heat flux and surface temperature of the vessel also are affected in the presence of decay heat.

  13. Irradiation effects on thermal properties of LWR hydride fuel

    Science.gov (United States)

    Terrani, Kurt; Balooch, Mehdi; Carpenter, David; Kohse, Gordon; Keiser, Dennis; Meyer, Mitchell; Olander, Donald

    2017-04-01

    Three hydride mini-fuel rods were fabricated and irradiated at the MIT nuclear reactor with a maximum burnup of 0.31% FIMA or ∼5 MWd/kgU equivalent oxide fuel burnup. Fuel rods consisted of uranium-zirconium hydride (U (30 wt%)ZrH1.6) pellets clad inside a LWR Zircaloy-2 tubing. The gap between the fuel and the cladding was filled with lead-bismuth eutectic alloy to eliminate the gas gap and the large temperature drop across it. Each mini-fuel rod was instrumented with two thermocouples with tips that are axially located halfway through the fuel centerline and cladding surface. In-pile temperature measurements enabled calculation of thermal conductivity in this fuel as a function of temperature and burnup. In-pile thermal conductivity at the beginning of test agreed well with out-of-pile measurements on unirradiated fuel and decreased rapidly with burnup.

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

    Energy Technology Data Exchange (ETDEWEB)

    Clayton, J C

    1987-10-01

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

  15. A New Reducing Regent: Dichloroindium Hydride

    Institute of Scientific and Technical Information of China (English)

    A. BABA; I. SHIBATA; N. HAYASHI

    2005-01-01

    @@ 1Introduction Among the hydride derivatives of group 13 elements, various types of aluminum hydrides and boron hydrides have been employed as powerful reduction tools. Indium hydrides have not received much attention,whereas the synthesis of indium trihydride (InH3) was reported several decades ago[1]. There have been no precedents for monometallic indium hydrides having practical reactivity, while activated hydrides such as an ate complex LiPhn InH4-n (n = 0- 2) and phosphine-coordinated indium hydrides readily reduce carbonyl compounds. In view of this background, we focused on the development of dichloroindium hydrides (Cl2InH) as novel reducing agents that bear characteristic features in both ionic and radical reactions.

  16. Hydride Olefin complexes of tantalum and niobium

    NARCIS (Netherlands)

    Klazinga, Aan Hendrik

    1979-01-01

    This thesis describes investigations on low-valent tantalum and niobium hydride and alkyl complexes, particularly the dicyclopentadienyl tantalum hydride olefin complexes Cp2Ta(H)L (L=olefin). ... Zie: Summary

  17. Complex and liquid hydrides for energy storage

    Energy Technology Data Exchange (ETDEWEB)

    Callini, Elsa; Atakli, Zuleyha Özlem Kocabas; Hauback, Bjørn C.; Orimo, Shin-ichi; Jensen, Craig; Dornheim, Martin; Grant, David; Cho, Young Whan; Chen, Ping; Hjörvarsson, Bjørgvin; de Jongh, Petra; Weidenthaler, Claudia; Baricco, Marcello; Paskevicius, Mark; Jensen, Torben R.; Bowden, Mark E.; Autrey, Thomas S.; Züttel, Andreas

    2016-03-10

    The research on complex hydrides for hydrogen storage was imitated by the discovery of Ti as a hydrogen sorption catalyst in NaAlH4 by Boris Bogdanovic in 1996. A large number of new complex hydride materials in various forms and combinations have been synthesized and characterized and the knowledge on the properties of complex hydrides and the synthesis methods has grown enormously since then. A significant part of the research groups active in the field of complex hydrides are collaborators in the IEA task 32. This paper reports about the important issues in the field of the complex hydride research, i.e. the synthesis of borohydrides, the thermodynamics of complex hydrides and their thermodynamic properties, the effects of size and confinement, the hydrogen sorption mechanism and the complex hydride composites as well as the properties of liquid complex hydrides. This paper is the result of the collaboration of several groups and excellent summary of the recent achievements.

  18. Luminescent properties of aluminum hydride

    Energy Technology Data Exchange (ETDEWEB)

    Baraban, A.P.; Gabis, I.E.; Dmitriev, V.A. [Saint-Petersburg State University, Department of Physics, Saint-Petersburg 198504 (Russian Federation); Dobrotvorskii, M.A., E-mail: mstislavd@gmail.com [Saint-Petersburg State University, Department of Physics, Saint-Petersburg 198504 (Russian Federation); Kuznetsov, V.G. [Saint-Petersburg State University, Department of Physics, Saint-Petersburg 198504 (Russian Federation); Matveeva, O.P. [National Mineral Resources University, Saint Petersburg 199106 (Russian Federation); Titov, S.A. [Petersburg State University of Railway Transport, Saint-Petersburg 190031 (Russian Federation); Voyt, A.P.; Elets, D.I. [Saint-Petersburg State University, Department of Physics, Saint-Petersburg 198504 (Russian Federation)

    2015-10-15

    We studied cathodoluminescence and photoluminescence of α-AlH{sub 3}– a likely candidate for use as possible hydrogen carrier in hydrogen-fueled vehicles. Luminescence properties of original α-AlH{sub 3} and α-AlH{sub 3} irradiated with ultraviolet were compared. The latter procedure leads to activation of thermal decomposition of α-AlH{sub 3} and thus has a practical implementation. We showed that the original and UV-modified aluminum hydride contain luminescence centers ‐ structural defects of the same type, presumably hydrogen vacancies, characterized by a single set of characteristic bands of radiation. The observed luminescence is the result of radiative intracenter relaxation of the luminescence center (hydrogen vacancy) excited by electrons or photons, and its intensity is defined by the concentration of vacancies, and the area of their possible excitation. UV-activation of the dehydrogenation process of aluminum hydride leads to changes in the spatial distribution of the luminescence centers. For short times of exposure their concentration increases mainly in the surface regions of the crystals. At high exposures, this process extends to the bulk of the aluminum hydride and ends with a decrease in concentration of luminescence centers in the surface region. - Highlights: • Aluminum hydride contains hydrogen vacancies which serve as luminescence centers. • The luminescence is the result of radiative relaxation of excited centers. • Hydride UV-irradiation alters distribution and concentration of luminescence centers.

  19. Photochemistry of Transition Metal Hydrides.

    Science.gov (United States)

    Perutz, Robin N; Procacci, Barbara

    2016-08-10

    Photochemical reactivity associated with metal-hydrogen bonds is widespread among metal hydride complexes and has played a critical part in opening up C-H bond activation. It has been exploited to design different types of photocatalytic reactions and to obtain NMR spectra of dilute solutions with a single pulse of an NMR spectrometer. Because photolysis can be performed on fast time scales and at low temperature, metal-hydride photochemistry has enabled determination of the molecular structure and rates of reaction of highly reactive intermediates. We identify five characteristic photoprocesses of metal monohydride complexes associated with the M-H bond, of which the most widespread are M-H homolysis and R-H reductive elimination. For metal dihydride complexes, the dominant photoprocess is reductive elimination of H2. Dihydrogen complexes typically lose H2 photochemically. The majority of photochemical reactions are likely to be dissociative, but hydride complexes may be designed with equilibrated excited states that undergo different photochemical reactions, including proton transfer or hydride transfer. The photochemical mechanisms of a few reactions have been analyzed by computational methods, including quantum dynamics. A section on specialist methods (time-resolved spectroscopy, matrix isolation, NMR, and computational methods) and a survey of transition metal hydride photochemistry organized by transition metal group complete the Review.

  20. Method of producing a chemical hydride

    Science.gov (United States)

    Klingler, Kerry M.; Zollinger, William T.; Wilding, Bruce M.; Bingham, Dennis N.; Wendt, Kraig M.

    2007-11-13

    A method of producing a chemical hydride is described and which includes selecting a composition having chemical bonds and which is capable of forming a chemical hydride; providing a source of a hydrocarbon; and reacting the composition with the source of the hydrocarbon to generate a chemical hydride.

  1. Hydrogen, lithium, and lithium hydride production

    Science.gov (United States)

    Brown, Sam W; Spencer, Larry S; Phillips, Michael R; Powell, G. Louis; Campbell, Peggy J

    2014-03-25

    A method of producing high purity lithium metal is provided, where gaseous-phase lithium metal is extracted from lithium hydride and condensed to form solid high purity lithium metal. The high purity lithium metal may be hydrided to provide high purity lithium hydride.

  2. Reactor Materials Research

    Energy Technology Data Exchange (ETDEWEB)

    Van Walle, E

    2001-04-01

    The activities of the Reactor Materials Research Department of the Belgian Nuclear Research Centre SCK-CEN in 2000 are summarised. The programmes within the department are focussed on studies concerning (1) fusion, in particular mechanical testing; (2) Irradiation Assisted Stress Corrosion Cracking (IASCC); (3) nuclear fuel; and (4) Reactor Pressure Vessel Steel (RPVS)

  3. R and D Developments. Research Programs on Irradiation Embrittlement of Reactor Vessel Steels; Programas de investigacion sobre fragilizacion por irradiacion de los aceros de vasija

    Energy Technology Data Exchange (ETDEWEB)

    Gomez Briceno, D.; Lapena, J.; Serrano, M.; Perosanz, F. [Ciemat. Madrid (Spain)

    2000-07-01

    Irradiation embrittlement of pressure vessel steels is a degradation mechanism time dependent that can lead to operational restrictions with adverse effects in the efficiency and life of a plant. For the last year, several research programs have been devoted to study the evaluation of neutronic radiation effect on mechanical properties of pressure vessel steels. However, at the present, there is a growing interest on the development of new methodologies to optimize the surveillance program information, and the understanding of the irradiation damage mechanism. This paper give an overview of international research programs, and on the R+D activities carried out by the Structural Materials Project on irradiation embrittlement on pressure vessel steels. (Author)

  4. Anodematerials for Metal Hydride Batteries

    DEFF Research Database (Denmark)

    Jensen, Jens Oluf

    1997-01-01

    This report describes the work on development of hydride forming alloys for use as electrode materials in metal hydride batteries. The work has primarily been concentrated on calcium based alloys derived from the compound CaNi5. This compound has a higher capacity compared with alloys used in today......’s hydride batteries, but a much poorer stability towards repeated charge/discharge cycling. The aim was to see if the cycleability of CaNi5 could be enhanced enough by modifications to make the compound a suitable electrode material. An alloying method based on mechanical alloying in a planetary ball mill...... by annealing at 700°C for 12 hours. The alloys appeared to be nanocrystalline with an average crystallite size around 10 nm before annealing. Special steel containers was developed for the annealing of the metal powders in inert atmosphere. The use of various annealing temperatures was investigated...

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1995-07-01

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

  6. Hydrogen Storage in Metal Hydrides

    Science.gov (United States)

    1990-08-01

    Hydrogen Storage Capacity Hydride by weight (%) [1) by volume (g/ml) [2] MgH2 7.00 0.101 Mg2NiH4 3.84 0,081 Mg2CuH4 2.04 - - 27 ...Include Security Classification) Hydrogen Storage in Metal Hydrides (U) 12. PERSONAL AUTHOR(S) DelaRosa, Mark J. 13a. TYPE OF REPORT 13b. TIME...objective of this program was to develop an economical process for pr-ducing a lightweight hydrogen storage medium by the chemical vapor infiltration

  7. Crystallography of shear transformations in zirconium hydrides

    Energy Technology Data Exchange (ETDEWEB)

    Cassidy, Michael Philip [Univ. of Illinois, Urbana-Champaign, IL (United States)

    1978-01-01

    The crystallography and substructure of the transformations which have been hypothesized as involving a martensitic shear, and which occur between zirconium hydrides were investigated. Specifically, the formation of gamma zirconium hydride from delta hydride and the delta hydride to epsilon hydride transformation were studied. The habit planes, orientation relationships, lattice invariant shears, and interface structures were determined by transmission electron microscopy and diffraction. Surface tilts were observed and measured with an interference microscope. The direction and magnitude of the shape strain produced by the formation of gamma were determined by the measurement of fiducial scratch displacements. These results were compared with the phenomenological crystallographic theory of martensitic transformations.

  8. Characteristics and Applications of Metal Hydrides

    Science.gov (United States)

    Egan, G. J.; Lynch, F. E.

    1987-01-01

    Report discusses engineering principles of uses of metal hydrides in spacecraft. Metal hydrides absorb, store, pump, compress, and expand hydrogen gas. Additionally, they release or absorb sizeable amounts of heat as they form and decompose - property adapted for thermal-energy management or for propulsion. Describes efforts to: Identify heat sources and sinks suitable for driving metal hydride thermal cycles in spacecraft; develop concepts for hydride subsystems employing available heating and cooling methods; and produce data base on estimated sizes, masses, and performances of hydride devices for spacecraft.

  9. ACCEPTABILITY ENVELOPE FOR METAL HYDRIDE-BASED HYDROGEN STORAGE SYSTEMS

    Energy Technology Data Exchange (ETDEWEB)

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

    2011-07-18

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

  10. Characterization of hydrides and delayed hydride cracking in zirconium alloys

    Science.gov (United States)

    Fang, Qiang

    This thesis tries to fill some of the missing gaps in the study of zirconium hydrides with state-of-art experiments, cutting edge tomographical technique, and a novel numerical algorithm. A new hydriding procedure is proposed. The new anode material and solution combination overcomes many drawbacks of the AECLRTM hydriding method and leads to superior hydriding result compared to the AECL RTM hydriding procedure. The DHC crack growth velocity of as-received Excel alloy and Zr-2.5Nb alloy together with several different heat treated Excel alloy samples are measured. While it already known that the DHC crack growth velocity increases with the increase of base metal strength, the finding that the transverse plane is the weaker plane for fatigue crack growth despite having higher resistance to DHC crack growth was unexpected. The morphologies of hydrides in a coarse grained Zircally-2 sample have been studied using synchrotron x-rays at ESRF with a new technique called Diffraction Contrast Tomography that uses simultaneous collection of tomographic data and diffraction data to determine the crystallographic orientation of crystallites (grains) in 3D. It has been previously limited to light metals such as Al or Mg (due to the use of low energy x-rays). Here we show the first DCT measurements using high energy x-rays (60 keV), allowing measurements in zirconium. A new algorithm of a computationally effcient way to characterize distributions of hydrides - in particular their orientation and/or connectivity - has been proposed. It is a modification of the standard Hough transform, which is an extension of the Hough transform widely used in the line detection of EBSD patterns. Finally, a basic model of hydrogen migration is built using ABAQUS RTM, which is a mature finite element package with tested modeling modules of a variety of physical laws. The coupling of hydrogen diffusion, lattice expansion, matrix deformation and phase transformation is investigated under

  11. Flaw distribution development from vessel ISI data

    Energy Technology Data Exchange (ETDEWEB)

    Foulds, J.R.; Kennedy, E.L. [Failure Analysis Associates, Inc., Menlo Park, CA (United States); Basin, S.L. [Joyce and Associates, Los Altos, CA (United States); Rosinski, S.T. [Sandia National Labs., Albuquerque, NM (United States)

    1991-12-31

    Previous attempts to develop flaw distributions for use in the structural integrity evaluation of pressurized water reactor (PWR) vessels have aimed at the estimation of a ``generic`` distribution applicable to all vessels. In contrast, this paper describes the analysis of vessel-specific in-service inspection (ISI) data for the development of a flaw distribution reliably representative of the condition of the particular vessel inspected. The application of the methodology may be extended to other vessels, but has been primarily developed for PWR reactor vessels. For this study, the flaw data analyzed included data obtained from three recently performed PWR vessel ISIs and from laboratory inspection of selected weldment sections of the Midland reactor vessel. The variability in both the character of the reviewed data (size range of flaws, number of flaws) and the UT (ultrasonic test) inspection system performance identified a need for analyzing the inspection results on a vessel-, or data set-specific basis. For this purpose, traditional histogram-based methods were inadequate, and a new methodology that can accept a very small number of flaws (typical of vessel-specific ISI results) and that includes consideration of inspection system flaw detection reliability, flaw sizing accuracy and flaw detection threshold, was developed. Results of the application of the methodology to each of the four PWR reactor vessel cases studied are presented and discussed.

  12. Flaw distribution development from vessel ISI data

    Energy Technology Data Exchange (ETDEWEB)

    Foulds, J.R.; Kennedy, E.L. (Failure Analysis Associates, Inc., Menlo Park, CA (United States)); Basin, S.L. (Joyce and Associates, Los Altos, CA (United States)); Rosinski, S.T. (Sandia National Labs., Albuquerque, NM (United States))

    1991-01-01

    Previous attempts to develop flaw distributions for use in the structural integrity evaluation of pressurized water reactor (PWR) vessels have aimed at the estimation of a generic'' distribution applicable to all vessels. In contrast, this paper describes the analysis of vessel-specific in-service inspection (ISI) data for the development of a flaw distribution reliably representative of the condition of the particular vessel inspected. The application of the methodology may be extended to other vessels, but has been primarily developed for PWR reactor vessels. For this study, the flaw data analyzed included data obtained from three recently performed PWR vessel ISIs and from laboratory inspection of selected weldment sections of the Midland reactor vessel. The variability in both the character of the reviewed data (size range of flaws, number of flaws) and the UT (ultrasonic test) inspection system performance identified a need for analyzing the inspection results on a vessel-, or data set-specific basis. For this purpose, traditional histogram-based methods were inadequate, and a new methodology that can accept a very small number of flaws (typical of vessel-specific ISI results) and that includes consideration of inspection system flaw detection reliability, flaw sizing accuracy and flaw detection threshold, was developed. Results of the application of the methodology to each of the four PWR reactor vessel cases studied are presented and discussed.

  13. Recommendations of the MRP-139: Inspection of Welds dissimilar in Nozzles PWR reactor vessel in Spain; Recomendaciones del MRP-139: Inspeccion de soldaduras disimilares en Vasijas de Reactor en Espana

    Energy Technology Data Exchange (ETDEWEB)

    Gadea, J. R.; Willke, A.; Regidor, J. J.; Tecnatom, S. A.

    2010-07-01

    The guide EPRI MRP-139, which provides the way forward for the inspection and evaluation of dissimilar butt welds, the primary system of PWR reactors, indicating the type of nondestructive testing to be done in these areas, based on discovered several cases of default in lnconel alloys 600 and 182 in American and European plants. The phenomenon of cracking.

  14. Simplified 3D model of a PWR reactor vessel using fluid dynamics code ANSYS CFX computational; Modelo simplificado 3D de la vasija de un reactor PWR mediante el codigo de dinamica de fluidos computacional ANSYS CFX

    Energy Technology Data Exchange (ETDEWEB)

    Martinez, M.; Miro, R.; Barrachina, T.; Verdu, G.

    2011-07-01

    This paper presents the results from the calculation of the steady state simulation with model of CFD (computational fluid dynamic) operating under conditions of operation at full power (Hot Full Power). Development and the CFD model results show the usefulness of these codes for calculating 3D of the variable thermohydraulics of these reactors.

  15. Research Progress on High-temperature Creep Behavior of Reactor Pressure Vessel%严重事故下反应堆压力容器材料高温蠕变研究进展

    Institute of Scientific and Technical Information of China (English)

    武志玮; 宁冬; 姚伟达

    2011-01-01

    介绍了近年来在假想堆芯熔化严重事故下国内外反应堆压力容器材料高温蠕变行为的研究进展及现状,着重阐述了在材料高温蠕变试验、缩比模型试验和数值模拟等方面取得的成果,并提出了目前存在的问题及未来的发展方向。%An overview of research status and progress on high-temperature creep behavior pressure vessel considering the hypothetical core melt down scenario is presented in this paper. is placed on accomplished achievements in The present problems as well as the future of reactor Emphasis creep tests, scale model experiments and numerical simulation. development are also suggested.

  16. Properties of nanoscale metal hydrides.

    Science.gov (United States)

    Fichtner, Maximilian

    2009-05-20

    Nanoscale hydride particles may exhibit chemical stabilities which differ from those of a macroscopic system. The stabilities are mainly influenced by a surface energy term which contains size-dependent values of the surface tension, the molar volume and an additional term which takes into account a potential reduction of the excess surface energy. Thus, the equilibrium of a nanoparticular hydride system may be shifted to the hydrogenated or to the dehydrogenated side, depending on the size and on the prefix of the surface energy term of the hydrogenated and dehydrogenated material. Additional complexity appears when solid-state reactions of complex hydrides are considered and phase segregation has to be taken into account. In such a case the reversibility of complex hydrides may be reduced if the nanoparticles are free standing on a surface. However, it may be enhanced if the system is enclosed by a nanoscale void which prevents the reaction partners on the dehydrogenated side from diffusing away from each other. Moreover, the generally enhanced diffusivity in nanocrystalline systems may lower the kinetic barriers for the material's transformation and, thus, facilitate hydrogen absorption and desorption.

  17. High gamma-rays irradiation tests of critical components for ITER (International Thermonuclear Experimental Reactor) in-vessel remote handling system

    Energy Technology Data Exchange (ETDEWEB)

    Obara, Kenjiro; Kakudate, Satoshi; Oka, Kiyoshi [Department of Fusion Engineering Research, Naka Fusion Research Establishment, Japan Atomic Energy Research Institute, Naka, Ibaraki (Japan)] [and others

    1999-02-01

    In ITER, the in-vessel remote handling is inevitably required to assemble and maintain the activated in-vessel components due to deuterium and tritium operation. Since the in-vessel remote handling system has to be operated under the intense of gamma ray irradiation, the components of the remote handling system are required to have radiation hardness so as to allow maintenance operation for a sufficient length of time under the ITER in-vessel environments. For this, the Japan, European and Russian Home Teams have extensively conducted gamma ray irradiation tests and quality improvements including optimization of material composition through ITER R and D program in order to develop radiation hard components which satisfy the doses from 10 MGy to 100 MGy at a dose rate of 1 x 10{sup 6} R/h (ITER R and D Task: T252). This report describes the latest status of radiation hard component development which has been conducted by the Japan Home Team in the ITER R and D program. The number of remote handling components tested is about seventy and these are categorized into robotics (Subtask 1), viewing system (Subtask 2) and common components (Subtask 3). The irradiation tests, including commercial base products for screening, modified products and newly developed products to improve the radiation hardness, were carried out using the gamma ray irradiation cells in Takasaki Establishment, JAERI. As a result, the development of the radiation hard components which can be tolerable for high temperature and gamma radiation has been well progressed, and many components, such as AC servo motor with ceramics insulated wire, optical periscope and CCD camera, have been newly developed. (author)

  18. Reactor Physics Programme

    Energy Technology Data Exchange (ETDEWEB)

    De Raedt, C

    2000-07-01

    The Reactor Physics and Department of SCK-CEN offers expertise in various areas of reactor physics, in particular in neutronics calculations, reactor dosimetry, reactor operation, reactor safety and control and non-destructive analysis on reactor fuel. This expertise is applied within the Reactor Physics and MYRRHA Research Department's own research projects in the VENUS critical facility, in the BR1 reactor and in the MYRRHA project (this project aims at designing a prototype Accelerator Driven System). Available expertise is also used in programmes external to the Department such as the reactor pressure steel vessel programme, the BR2 reactor dosimetry, and the preparation and interpretation of irradiation experiments. Progress and achievements in 1999 in the following areas are reported on: (1) investigations on the use of military plutonium in commercial power reactors; (2) neutron and gamma calculations performed for BR-2 and for other reactors; (3) the updating of neutron and gamma cross-section libraries; (4) the implementation of reactor codes; (6) the management of the UNIX workstations; and (6) fuel cycle studies.

  19. A study on the fusion reactor - Development of electrical insulation coating processes for vacuum vessel components of KT-2 tokamak by plasma spray techniques

    Energy Technology Data Exchange (ETDEWEB)

    Hong, Sang Hee; Choi, Byung Yong; Ahn, Hyun; Ju, Won Tae; Eom, You Sub [Seoul National University, Seoul (Korea, Republic of)

    1995-08-01

    For the fabrication of insulation coatings with good vacuum tightness, mechanical and electrical properties needed for voltage breaker and plasma facing components of tokamak vacuum vessel, a plasma spraying system equipped= with an improved power supply and a precision powder feeder is employed for the development of the optimum processes for ceramic insulation coatings. The material properties of the ceramic coatings for tokamak vacuum vessel components are evaluated by material tests and analyses to determine optimum processing parameters for insulation coatings. As a result of material evaluation for Al{sub 2}O{sub 3} and Al{sub 2}O{sub 3}-TiO{sub 2} ceramic insulation coatings fabricated, Al{sub 2}O{sub 3}-3%TiO{sub 2} ceramic turn out to be the best insulation coating for tokamak use in respect of electrical and mechanical properties. Al{sub 2}O{sub 3} coating with dielectric strength values of more than 26 kV/mm can also be applicable to tokamak vacuum vessel components for electric insulation by improving its low adhesive strength. 23 refs., 9 tabs., 14 figs. (author)

  20. Research vessels

    Digital Repository Service at National Institute of Oceanography (India)

    Rao, P.S.

    The role of the research vessels as a tool for marine research and exploration is very important. Technical requirements of a suitable vessel and the laboratories needed on board are discussed. The history and the research work carried out...

  1. Air passivation of metal hydride beds for waste disposal

    Energy Technology Data Exchange (ETDEWEB)

    Klein, J. E.; Hsu, R. H. [Savannah River National Laboratory, Aiken, SC 29808 (United States)

    2008-07-15

    One waste acceptance criteria for hydride bed waste disposal is that the bed be non-pyrophoric. Batch-wise air ingress tests were performed which determined the amount of air consumed by a metal hydride bed. A desorbed, 4.4 kg titanium prototype hydride storage vessel (HSV) produced a 4.4 deg.C internal temperature rise upon the first air exposure cycle and a 0.1 deg.C temperature rise upon a second air exposure. A total of 346 sec air was consumed by the bed (0.08 sec per gram Ti). A desorbed, 9.66 kg LaNi{sub 4.25}Al{sub 0.75} prototype storage bed experienced larger temperature rises over successive cycles of air ingress and evacuation. The cycles were performed over a period of days with the bed effectively passivated after the 12. cycle. Nine to ten STP-L of air reacted with the bed producing both oxidized metal and water. (authors)

  2. Nuclear Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Hogerton, John

    1964-01-01

    This pamphlet describes how reactors work; discusses reactor design; describes research, teaching, and materials testing reactors; production reactors; reactors for electric power generation; reactors for supply heat; reactors for propulsion; reactors for space; reactor safety; and reactors of tomorrow. The appendix discusses characteristics of U.S. civilian power reactor concepts and lists some of the U.S. reactor power projects, with location, type, capacity, owner, and startup date.

  3. Evaluation of defects induced by neutron radiation in reactor pressure vessels steels; Evaluacion de los defectos inducidos por la radiacion neutronica en los aceros de vasijas

    Energy Technology Data Exchange (ETDEWEB)

    Lopez Jimenez, J.

    1978-07-01

    We have developed a method for calculating the production of neutron induced defects (depleted zone and crowdions) in ferritic pressure vessel steels for different neutron spectra. They have been analysed both the recoil primary atoms produced by elastic and inelastic collisions with fast neutrons and the ones produced by gamma-ray emission by thermal neutron absorption. Theoretical modelling of increasing in the ductile-brittle transition temperature of ferritic steels has been correlated with experimental data at irradiation temperature up to 400 degree centigree (Author) 15 refs.

  4. Strain evolution during hydride precipitation in Zircaloy-4 observed with synchrotron X-ray diffraction

    Science.gov (United States)

    Blackmur, M. S.; Preuss, M.; Robson, J. D.; Zanellato, O.; Cernik, R. J.; Ribeiro, F.; Andrieux, J.

    2016-06-01

    Synchrotron X-ray diffraction was used to evaluate strain evolution observed in Zircaloy-4 undergoing hydride precipitation during a range of thermal operations. During continuous heating, a change in the constraining effect of the matrix was observed at a temperature of 280 °C, thought to be the result of matrix dilatation from interstitial hydrogen. A deconvolution of the thermal, chemical and mechanical sources of strain during quench and dwell operations identified a non-negligible mechanical effect in the matrix. During these dwells, slow strain rate relaxation of elastic strains was seen in the matrix and hydride, suggesting that time dependent relaxation of misfit stresses may be possible at reactor relevant temperatures. Notable anisotropy was observed between the rolling and transverse directions, identified as being the likely product of a similar anisotropy in the relaxation of the hydride misfit between the α and α matrix directions, owing to the differing coherency of these two interfaces.

  5. Hydride-induced embrittlement of Zircaloy-4 cladding under plane-strain tension

    Science.gov (United States)

    Daum, Robert S.

    The mechanical response of high-burnup Zircaloy-4 fuel cladding subjected to a postulated reactivity initiated accident (referred to as a rod ejection accident (REA) in a pressurized water reactor) can be affected by hydrogen embrittlement. This study addresses the hydrogen embrittlement of non-irradiated, stress-relieved Zircaloy-4 cladding under conditions (state of stress and temperature) relevant to those of a reactivity initiated accident. Specifically, the study has investigated the effects of a concentrated density of hydride particles (in the form of a rim at the outer surface of the cladding tube introduced by gas-charging) on the cladding ductility when tested under a near-plane-strain tension at 25, 300, and 375°C. The influence of the hydride-rim thickness and local hydrogen contents on cladding ductility is studied as a function of temperature and correlated with the hydride microstructure. Using synchrotron x-ray diffraction, this study has found that the delta-hydride phase (i.e., ZrHx, where x ≈ 1.66) is the predominant hydride phase to precipitate in stress-relieved Zircaloy-4 cladding for hydrogen contents up to 1250 wt ppm. At hydrogen contents above 2700 wt ppm, although delta-hydride is still the majority phase, both gamma- and epsilon-hydride phases are also observed. The volume fraction of hydrides was estimated as a function of hydrogen content, using the diffracted x-ray intensities. These estimated values agree well with calculated values assuming hydride precipitates are delta-hydride. Under near-plane-strain hoop tension, the ductility and fracture of the cladding is highly dependent on both the hydride-rim thickness and the testing temperature. At room temperature, due to a high density of hydride particles within the rim, a Mode I crack is injected shortly after yielding. This limits cladding ductility, such that it decreases with increasing thickness of the hydride rim. Cladding containing hydride rims with a thickness of ≥100

  6. Understanding of hydriding mechanisms of zircaloy-4 alloy during corrosion in PWR simulated conditions and influence of zirconium hydrides on zircaloy-4 corrosion

    Energy Technology Data Exchange (ETDEWEB)

    Bisor-Melloul, C.; Tupin, M.; Bossis, P. [DEN/DANS/DMN/SEMI, CEA/Saclay, 91191 Gif-sur-Yvette (France); Chene, J. [DEN/DANS/DPC/SCCME, CEA/Saclay, 91191 Gif-sur-Yvette (France); Bechade, J.L. [DEN/DANS/DMN/SRMA, CEA/Saclay, 91191 Gif-sur-Yvette (France); Motta, A. [Mechanical and Nuclear Engineering Department, Penn State University, 227 Reber Building, University Park, PA 16802 (United States)

    2010-07-01

    Zirconium alloys are widely used as fuel claddings in Power Water Reactors. As they represent the first containment barrier to fission products, their mechanical integrity is essential for nuclear safety. During their corrosion in primary water, some of the hydrogen involved in the oxidation reaction with water ingresses into the alloy through the oxide layer. In the metallic matrix, once the solid solution limit is reached at the irradiation temperature, hydrogen precipitates as Zr hydrides mainly located just under the metal/oxide interface due to the thermal gradient across the cladding. As these hydrides may contribute to a larger oxide thickness and to a more fragile behaviour of the cladding, the minimization of hydrogen pick-up is required. Accordingly, since the Zircaloy-4 (Zr-1.3Sn-0.2Fe-0.1Cr) alloy is known to be sensitive to this phenomenon, the understanding of its hydriding mechanism and of the influence of zirconium hydrides on its corrosion behaviour is needed. Regarding the study of the hydriding mechanism, isotopic exchanges were carried out in D{sub 2}O environment at 360 deg. C and led to the localization, in the oxide scales, of the limiting step for the hydrogen diffusion. To estimate an apparent diffusion coefficient of hydrogen in the oxide formed on Zircaloy-4, we firstly based on SIMS profiles and penetration depth of deuterium in the dense part of the oxide film. Secondly, ERDA estimation of the hydrogen content in zirconia and fusion measurements of the hydrogen content in both metal and oxide were used to estimate a hydrogen flux absorbed by the alloy and hence to deduce an apparent diffusion coefficient. Finally, these two methods lead to quite similar values (between 2.10{sup -14} cm{sup 2}/s and 6.10{sup -14} cm{sup 2}/s) which are in accordance with bibliography. Concerning the impact of hydrides on the corrosion of Zircaloy-4, several pre-hydrided and reference samples were corroded simultaneously in primary water at 360 deg. C

  7. The renaissance of hydrides as energy materials

    Science.gov (United States)

    Mohtadi, Rana; Orimo, Shin-Ichi

    2017-02-01

    Materials based on hydrides have been the linchpin in the development of several practical energy storage technologies, of which the most prominent example is nickel-metal hydride batteries. Motivated by the need to meet the future's energy demand, the past decade has witnessed substantial advancements in the research and development of hydrides as media for hydrogen energy storage. More recently, new and rapidly evolving discoveries have positioned hydrides as highly promising materials for future electrochemical energy storage, such as electrolytes for mono- and divalent batteries, and anodes for lithium-ion batteries. In addition, the potential of hydrides in efficient power transmission has been recently revealed. In this Review, we highlight key advances and illustrate how the versatility of hydrides has not only yielded a meaningful past, but also ensures a very bright future.

  8. Rechargeable metal hydrides for spacecraft application

    Science.gov (United States)

    Perry, J. L.

    1988-01-01

    Storing hydrogen on board the Space Station presents both safety and logistics problems. Conventional storage using pressurized bottles requires large masses, pressures, and volumes to handle the hydrogen to be used in experiments in the U.S. Laboratory Module and residual hydrogen generated by the ECLSS. Rechargeable metal hydrides may be competitive with conventional storage techniques. The basic theory of hydride behavior is presented and the engineering properties of LaNi5 are discussed to gain a clear understanding of the potential of metal hydrides for handling spacecraft hydrogen resources. Applications to Space Station and the safety of metal hydrides are presented and compared to conventional hydride storage. This comparison indicates that metal hydrides may be safer and require lower pressures, less volume, and less mass to store an equivalent mass of hydrogen.

  9. Decontamination of the reactor pressure vessel and further internals and auxiliary systems in the German boiling water reactor Isar-1; Dekontamination des RDB inkl. der Einbauten wie Dampftrockner und Wasserabscheider sowie der angeschlossenen Hilfssysteme im deutschen Siedewasserreaktor ISAR 1

    Energy Technology Data Exchange (ETDEWEB)

    Fischer, Michael; Sempere Belda, Luis; Basu, Ashim; Topf, Christian [AREVA GmbH, Erlangen (Germany). Abt. Chemistry Services; Erbacher, Thomas; Hiermer, Thomas; Schnurr, Bernhard; Appeldorn, Thomas van [E.ON Kernkraft GmbH, Kernkraftwerk ISAR, Essenbach (Germany). Abt. Maschinentechnik; Volkmann, Christian [ESG Engineering Services GmbH, Greifswald (Germany)

    2015-12-15

    The German nuclear power plant ISAR 1 (KKI 1), a 878 MWe boiling water reactor of KWU design, was shut down on March 17{sup th}, 2011. With the objective to minimize the plants activity inventory accompanied by the reduction of contact dose rates of systems and components the project 'decontamination of the RPV incl. steam dryer and water separator and the connected auxiliary systems' was implemented in the first quarter of 2015. One major focus within the project was the specific in-situ decontamination of the steam dryer.

  10. Nanostructured, complex hydride systems for hydrogen generation

    Directory of Open Access Journals (Sweden)

    Robert A. Varin

    2015-02-01

    Full Text Available Complex hydride systems for hydrogen (H2 generation for supplying fuel cells are being reviewed. In the first group, the hydride systems that are capable of generating H2 through a mechanical dehydrogenation phenomenon at the ambient temperature are discussed. There are few quite diverse systems in this group such as lithium alanate (LiAlH4 with the following additives: nanoiron (n-Fe, lithium amide (LiNH2 (a hydride/hydride system and manganese chloride MnCl2 (a hydride/halide system. Another hydride/hydride system consists of lithium amide (LiNH2 and magnesium hydride (MgH2, and finally, there is a LiBH4-FeCl2 (hydride/halide system. These hydride systems are capable of releasing from ~4 to 7 wt.% H2 at the ambient temperature during a reasonably short duration of ball milling. The second group encompasses systems that generate H2 at slightly elevated temperature (up to 100 °C. In this group lithium alanate (LiAlH4 ball milled with the nano-Fe and nano-TiN/TiC/ZrC additives is a prominent system that can relatively quickly generate up to 7 wt.% H2 at 100 °C. The other hydride is manganese borohydride (Mn(BH42 obtained by mechano-chemical activation synthesis (MCAS. In a ball milled (2LiBH4 + MnCl2 nanocomposite, Mn(BH42 co-existing with LiCl can desorb ~4.5 wt.% H2 at 100 °C within a reasonable duration of dehydrogenation. Practical application aspects of hydride systems for H2 generation/storage are also briefly discussed.

  11. Use of reversible hydrides for hydrogen storage

    Science.gov (United States)

    Darriet, B.; Pezat, M.; Hagenmuller, P.

    1980-01-01

    The addition of metals or alloys whose hydrides have a high dissociation pressure allows a considerable increase in the hydrogenation rate of magnesium. The influence of temperature and hydrogen pressure on the reaction rate were studied. Results concerning the hydriding of magnesium rich alloys such as Mg2Ca, La2Mg17 and CeMg12 are presented. The hydriding mechanism of La2Mg17 and CeMg12 alloys is given.

  12. Inhibited solid propellant composition containing beryllium hydride

    Science.gov (United States)

    Thompson, W. W. (Inventor)

    1978-01-01

    An object of this invention is to provide a composition of beryllium hydride and carboxy-terminated polybutadiene which is stable. Another object of this invention is to provide a method for inhibiting the reactivity of beryllium hydride toward carboxy-terminated polybutadiene. It was found that a small amount of lecithin inhibits the reaction of beryllium hydride with the acid groups in carboxy terminated polybutadiene.

  13. Anodematerials for Metal Hydride Batteries

    DEFF Research Database (Denmark)

    Jensen, Jens Oluf

    1997-01-01

    by annealing at 700°C for 12 hours. The alloys appeared to be nanocrystalline with an average crystallite size around 10 nm before annealing. Special steel containers was developed for the annealing of the metal powders in inert atmosphere. The use of various annealing temperatures was investigated......This report describes the work on development of hydride forming alloys for use as electrode materials in metal hydride batteries. The work has primarily been concentrated on calcium based alloys derived from the compound CaNi5. This compound has a higher capacity compared with alloys used in today...... was developed. The parameters milling time, milling intensity, number of balls and form of the alloying metals were investigated. Based on this a final alloying technique for the subsequent preparation of electrode materials was established. The technique comprises milling for 4 hours twice possibly followed...

  14. Model test of reactor vessel wall. Pt. 2. Test performance, measurement and partial evaluation; Modellkoerperversuch zur Reaktortankwand. T. 2; Versuchsdurchfuehrung, Messung und Teilauswertung

    Energy Technology Data Exchange (ETDEWEB)

    Maile, K.; Eckert, W.; Theofel, H.; Purper, H.

    1992-07-01

    Due to test interruption because of cut promotion means, the original objective of the project - verification of reactor wall design - could not be achieved because by that point in time the slabs had not yet failed (DIN 1.4948 = X 6 CrNi 18 11). Considering, however, the elongation curve, in particular that of the faulty slab, failure at an earlier stress cycle value than calculated is highly probable. (orig./HP) [Deutsch] Aufgrund des foerderungsbedingten Abbruchs der Versuche konnte die urspruengliche Zielsetzung des Vorhabens - Verifizierung der Auslegung der Reaktorwand - nicht erreicht werden, da ein Versagen der Platten bis zu diesem Zeitpunkt noch nicht vorlag (DIN 1.4948 = X 6 CrNi 18 11). Betrachtet man jedoch den Dehnungsverlauf insbesondere in der Platte mit Fehlern ist zu vermuten, dass ein Versagen zu einer frueheren Lastspielzahl als berechnet sehr wahrscheinlich ist. (orig./HP)

  15. Predicting formation enthalpies of metal hydrides

    DEFF Research Database (Denmark)

    Andreasen, A.

    2004-01-01

    In order for the hydrogen based society viz. a society in which hydrogen is the primary energy carrier to become realizable an efficient way of storing hydrogen is required. For this purpose metal hydrides are serious candidates. Metal hydrides are formedby chemical reaction between hydrogen...... and metal and for the stable hydrides this is associated with release of heat (#DELTA#H_f ). The more thermodynamically stable the hydride, the larger DHf, and the higher temperature is needed in order to desorphydrogen (reverse reaction) and vice versa. For practical application the temperature needed...

  16. Research on Metal Hydride Compressor System

    Institute of Scientific and Technical Information of China (English)

    1999-01-01

    Ti-Zr series Laves phase hydrogen storage alloys with good hydrogen storage properties, such as large hydrogen capacity, rapid hydriding and dehydriding rate, high compression ratio, gentle plateau, small hysteresis, easily being activated and long cyclic stability etc. for metal hydride compressor have been investigated. In addition, a hydride compressor with special characteristics, namely, advanced filling method, good heat transfer effect and reasonable structural design etc. has also been constructed. A hydride compressor cryogenic system has been assembled coupling the compressor with a J-T micro-throttling refrigeration device and its cooling capacity can reach 0.4 W at 25 K.

  17. Thermodynamic Hydricity of Transition Metal Hydrides.

    Science.gov (United States)

    Wiedner, Eric S; Chambers, Matthew B; Pitman, Catherine L; Bullock, R Morris; Miller, Alexander J M; Appel, Aaron M

    2016-08-10

    Transition metal hydrides play a critical role in stoichiometric and catalytic transformations. Knowledge of free energies for cleaving metal hydride bonds enables the prediction of chemical reactivity, such as for the bond-forming and bond-breaking events that occur in a catalytic reaction. Thermodynamic hydricity is the free energy required to cleave an M-H bond to generate a hydride ion (H(-)). Three primary methods have been developed for hydricity determination: the hydride transfer method establishes hydride transfer equilibrium with a hydride donor/acceptor pair of known hydricity, the H2 heterolysis method involves measuring the equilibrium of heterolytic cleavage of H2 in the presence of a base, and the potential-pKa method considers stepwise transfer of a proton and two electrons to give a net hydride transfer. Using these methods, over 100 thermodynamic hydricity values for transition metal hydrides have been determined in acetonitrile or water. In acetonitrile, the hydricity of metal hydrides spans a range of more than 50 kcal/mol. Methods for using hydricity values to predict chemical reactivity are also discussed, including organic transformations, the reduction of CO2, and the production and oxidation of hydrogen.

  18. Coinage Metal Hydrides: Synthesis, Characterization, and Reactivity.

    Science.gov (United States)

    Jordan, Abraham J; Lalic, Gojko; Sadighi, Joseph P

    2016-08-10

    Hydride complexes of copper, silver, and gold encompass a broad array of structures, and their distinctive reactivity has enabled dramatic recent advances in synthesis and catalysis. This Review summarizes the synthesis, characterization, and key stoichiometric reactions of isolable or observable coinage metal hydrides. It discusses catalytic processes in which coinage metal hydrides are known or probable intermediates, and presents mechanistic studies of selected catalytic reactions. The purpose of this Review is to convey how developments in coinage metal hydride chemistry have led to new organic transformations, and how developments in catalysis have in turn inspired the synthesis of reactive new complexes.

  19. Crystal structure of gold hydride

    Energy Technology Data Exchange (ETDEWEB)

    Degtyareva, Valentina F., E-mail: degtyar@issp.ac.ru

    2015-10-05

    Highlights: • Volume expansion of metal hydrides is due to the increase in the s-band filling. • AuH structure is similar to that of Hg having one more s electron compared to Au. • Structure stability of both Hg and AuH is governed by the Hume-Rothery rule. - Abstract: A number of transition metal hydrides with close-packed metal sublattices of fcc or hcp structures with hydrogen in octahedral interstitial positions were obtained by the high-pressure-hydrogen technique described by Ponyatovskii et al. (1982). In this paper we consider volume increase of metals by hydrogenation and possible crystal structure of gold hydride in relation with the structure of mercury, the nearest neighbor of Au in the Periodic table. Suggested structure of AuH has a basic tetragonal body-centered cell that is very similar to the mercury structure Hg-t I 2. The reasons of stability for this structure are discussed within the model of Fermi sphere–Brillouin zone interactions.

  20. Fundamental experiments on hydride reorientation in zircaloy

    Science.gov (United States)

    Colas, Kimberly B.

    In the current study, an in-situ X-ray diffraction technique using synchrotron radiation was used to follow directly the kinetics of hydride dissolution and precipitation during thermomechanical cycles. This technique was combined with conventional microscopy (optical, SEM and TEM) to gain an overall understanding of the process of hydride reorientation. Thus this part of the study emphasized the time-dependent nature of the process, studying large volume of hydrides in the material. In addition, a micro-diffraction technique was also used to study the spatial distribution of hydrides near stress concentrations. This part of the study emphasized the spatial variation of hydride characteristics such as strain and morphology. Hydrided samples in the shape of tensile dog-bones were used in the time-dependent part of the study. Compact tension specimens were used during the spatial dependence part of the study. The hydride elastic strains from peak shift and size and strain broadening were studied as a function of time for precipitating hydrides. The hydrides precipitate in a very compressed state of stress, as measured by the shift in lattice spacing. As precipitation proceeds the average shift decreases, indicating average stress is reduced, likely due to plastic deformation and morphology changes. When nucleation ends the hydrides follow the zirconium matrix thermal contraction. When stress is applied below the threshold stress for reorientation, hydrides first nucleate in a very compressed state similar to that of unstressed hydrides. After reducing the average strain similarly to unstressed hydrides, the average hydride strain reaches a constant value during cool-down to room temperature. This could be due to a greater ease of deforming the matrix due to the applied far-field strain which would compensate for the strains due to thermal contraction. Finally when hydrides reorient, the average hydride strains become tensile during the first precipitation regime and

  1. Metal Hydrides for Rechargeable Batteries

    Energy Technology Data Exchange (ETDEWEB)

    Valoeen, Lars Ole

    2000-03-01

    Rechargeable battery systems are paramount in the power supply of modern electronic and electromechanical equipment. For the time being, the most promising secondary battery systems for the future are the lithium-ion and the nickel metal hydride (NiMH) batteries. In this thesis, metal hydrides and their properties are described with the aim of characterizing and improving those. The thesis has a special focus on the AB{sub 5} type hydrogen storage alloys, where A is a rare earth metal like lanthanum, or more commonly misch metal, which is a mixture of rare earth metals, mainly lanthanum, cerium, neodymium and praseodymium. B is a transition metal, mainly nickel, commonly with additions of aluminium, cobalt, and manganese. The misch metal composition was found to be very important for the geometry of the unit cell in AB{sub 5} type alloys, and consequently the equilibrium pressure of hydrogen in these types of alloys. The A site substitution of lanthanum by misch metal did not decrease the surface catalytic properties of AB{sub 5} type alloys. B-site substitution of nickel with other transition elements, however, substantially reduced the catalytic activity of the alloy. If the internal pressure within the electrochemical test cell was increased using inert argon gas, a considerable increase in the high rate charge/discharge performance of LaNi{sub 5} was observed. An increased internal pressure would enable the utilisation of alloys with a high hydrogen equivalent pressure in batteries. Such alloys often have favourable kinetics and high hydrogen diffusion rates and thus have a potential for improving the high current discharge rates in metal hydride batteries. The kinetic properties of metal hydride electrodes were found to improve throughout their lifetime. The activation properties were found highly dependent on the charge/discharge current. Fewer charge/discharge cycles were needed to activate the electrodes if a small current was used instead of a higher

  2. Metal Hydrides for Rechargeable Batteries

    Energy Technology Data Exchange (ETDEWEB)

    Valoeen, Lars Ole

    2000-03-01

    Rechargeable battery systems are paramount in the power supply of modern electronic and electromechanical equipment. For the time being, the most promising secondary battery systems for the future are the lithium-ion and the nickel metal hydride (NiMH) batteries. In this thesis, metal hydrides and their properties are described with the aim of characterizing and improving those. The thesis has a special focus on the AB{sub 5} type hydrogen storage alloys, where A is a rare earth metal like lanthanum, or more commonly misch metal, which is a mixture of rare earth metals, mainly lanthanum, cerium, neodymium and praseodymium. B is a transition metal, mainly nickel, commonly with additions of aluminium, cobalt, and manganese. The misch metal composition was found to be very important for the geometry of the unit cell in AB{sub 5} type alloys, and consequently the equilibrium pressure of hydrogen in these types of alloys. The A site substitution of lanthanum by misch metal did not decrease the surface catalytic properties of AB{sub 5} type alloys. B-site substitution of nickel with other transition elements, however, substantially reduced the catalytic activity of the alloy. If the internal pressure within the electrochemical test cell was increased using inert argon gas, a considerable increase in the high rate charge/discharge performance of LaNi{sub 5} was observed. An increased internal pressure would enable the utilisation of alloys with a high hydrogen equivalent pressure in batteries. Such alloys often have favourable kinetics and high hydrogen diffusion rates and thus have a potential for improving the high current discharge rates in metal hydride batteries. The kinetic properties of metal hydride electrodes were found to improve throughout their lifetime. The activation properties were found highly dependent on the charge/discharge current. Fewer charge/discharge cycles were needed to activate the electrodes if a small current was used instead of a higher

  3. Evaluation of In-Vessel Corium Retention under a Severe Accident

    Energy Technology Data Exchange (ETDEWEB)

    Park, Rae-Joon; Kang, Kyung-Ho; Ha, Kwang-Soon; Kim, Jong-Tae; Koo, Kil-Mo; Cho, Young-Ro; Hong, Seong-Wan; Kim, Sang-Baik; Kim, Hee-Dong

    2008-02-15

    The current study on In-Vessel corium Retention and its application activities to the actual nuclear power plant have been reviewed and discussed in this study. Severe accident sequence which determines an initial condition of the IVR has been evaluated and late phase melt progression, heat transfer on the outer reactor vessel, and in-vessel corium cooling mechanism have been estimated in detail. During the high pressure sequence of the reactor coolant system, a natural circulation flow of the hot steam leads to a failure of the pressurizer surge line before the reactor vessel failure, which leads to a rapid decrease of the reactor coolant system pressure. The results of RASPLAV/MASCA study by OECD/NEA have shown that a melt stratification has occurred in the lower plenum of the reactor vessel. In particular, laver inversion has occurred, which is that a high density of the metal melt moves to the lower part of the oxidic melt layer. A method of heat transfer enhancement on the outer reactor vessel is an optimal design of the reactor vessel insulation for an increase of the natural circulation flow between the outer reactor vessel and the its insulation, and an increase of the critical Heat flux on the outer reactor vessel by using various method, such as Nono fluid, coated reactor vessel, and so on. An increase method of the in-vessel melt cooling is a development of the In-vessel core catcher and a decrease of focusing effect in the metal layer.

  4. Kinetics of hydride front in Zircaloy-2 and H release from a fractional hydrided surface

    Energy Technology Data Exchange (ETDEWEB)

    Diaz, M.; Gonzalez-Gonzalez, A.; Moya, J. S.; Remartinez, B.; Perez, S.; Sacedon, J. L. [Instituto de Ciencia de Materiales de Madrid (CSIC), Sor Juana Ines de la Cruz 3, Cantoblanco, 28049 Madrid (Spain); Iberdrola, Tomas Redondo 3, 28033 Madrid (Spain); Instituto de Ciencia de Materiales de Madrid (CSIC), Sor Juana Ines de la Cruz 3, Cantoblanco, 28049 Madrid (Spain)

    2009-07-15

    The authors study the hydriding process on commercial nuclear fuel claddings from their inner surface using an ultrahigh vacuum method. The method allows determining the incubation and failure times of the fuel claddings, as well as the dissipated energy and the partial pressure of the desorbed H{sub 2} from the outer surface of fuel claddings during the hydriding process. The correlation between the hydriding dissipated energy and the amount of zirconium hydride (formed at different stages of the hydriding process) leads to a near t{sup 1/2} potential law corresponding to the time scaling of the reaction for the majority of the tested samples. The calibrated relation between energy and hydride thickness allows one to calculate the enthalpy of the {delta}-ZrH{sub 1.5} phase. The measured H{sub 2} desorption from the external surface is in agreement with a proposed kinetic desorption model from the hydrides precipitated at the surface.

  5. Quantification of brittle-ductile failure behavior of ferritic reactor pressure vessel steels using the Small-Punch-Test and micromechanical damage models

    Energy Technology Data Exchange (ETDEWEB)

    Linse, T., E-mail: thomas.linse@tu-dresden.de [Technische Universität Dresden, Institute for Solid Mechanics, 01062 Dresden (Germany); Kuna, M. [Technische Universität Bergakademie Freiberg, Institute of Mechanics and Fluid Dynamics, Lampadiusstrasse 4, 09596 Freiberg (Germany); Viehrig, H.-W. [Helmholtz-Zentrum Dresden-Rossendorf, P.O. Box 510119, 01328 Dresden (Germany)

    2014-09-22

    Two German ferritic pressure vessel steels are examined in the brittle to ductile transition regime as a function of temperature and irradiation. The experiments are done by a miniaturized Small-Punch-Test in hot cells within the temperature range of −185 °C up to 70 °C. From the load–displacement curve of the SPT, the yield curves and parameters of both a non-local GURSON-TVERGAARD-NEEDLEMAN ductile damage model and a modified BEREMIN model are identified. The influence of temperature and irradiation on the model parameters is analyzed. All parameters are verified by comparison with results from standard test methods. The parameters, identified from SPT, are used to simulate the failure behavior in standard fracture mechanics specimens. In the upper shelf, the non-local GTN-model is applied to simulate crack resistance curves, from where the fracture toughness data could be successfully predicted. In the lower shelf, the WEIBULL-stress of the specimens was computed to find out the statistics of fracture toughness values. Finally, the modified BEREMIN model and the non-local ductile damage model were combined to evaluate the failure of fracture specimens in the brittle-ductile transition region. This way, an acceptable agreement with Master-curve data for non-irradiated steels could be achieved in the whole temperature range.

  6. Ionic conduction of lithium hydride single crystals

    Energy Technology Data Exchange (ETDEWEB)

    Pilipenko, G.I.; Oparin, D.V.; Zhuravlev, N.A.; Gavrilov, F.F.

    1987-09-01

    Using the electrical-conductivity- and NMR-measurement- methods, the ionic-conduction mechanism is established in stoichiometric lithium hydride single crystals. The activation energies of migration of anion- and cation-vacancies and the formation of Schottky-pair defects are determined. They assume that the mechanisms of self-diffusion and conductivity are different in lithium hydride.

  7. Visual interface for the automation of the instrumented pendulum of Charpy tests used in the surveillance program of reactors vessel of nuclear power plants; Interfase visual para la automatizacion del pendulo instrumentado de pruebas Charpy utilizado en el programa de vigilancia de la vasija de reactores de centrales nucleares

    Energy Technology Data Exchange (ETDEWEB)

    Rojas S, A.S.; Sainz M, E.; Ruiz E, J.A. [ININ, Carretera Mexico-Toluca Km.36.5, Mpio. de Ocoyoacac, Estado de Mexico (Mexico)]. E-mail: asrs@nuclear.inin.mx; esm@nuclear.inin.mx; jare@nuclear.inin.mx

    2004-07-01

    Inside the Programs of Surveillance of the nuclear power stations periodic information is required on the state that keep the materials with those that builds the vessel of the reactor. This information is obtained through some samples or test tubes that are introduced inside the core of the reactor and it is observed if its physical characteristics remain after having been subjected to the radiation changes and temperature. The rehearsal with the instrumented Charpy pendulum offers information on the behavior of fracture dynamics of a material. In the National Institute of Nuclear Research (ININ) it has an instrumented Charpy pendulum. The operation of this instrument is manual, having inconveniences to carry out rehearsals with radioactive material, handling of high and low temperatures, to fulfill the normative ones for the realization of the rehearsals, etc. In this work the development of a computational program is presented (virtual instrument), for the automation of the instrumented pendulum. The system has modules like: Card of data acquisition, signal processing, positioning system, tempered system, pneumatic system, compute programs like it is the visual interface for the operation of the instrumented Charpy pendulum and the acquisition of impact signals. This system shows that given the characteristics of the nuclear industry with radioactive environments, the virtual instrumentation and the automation of processes can contribute to diminish the risks to the personnel occupationally exposed. (Author)

  8. Submillimeter Spectroscopy of Hydride Molecules

    Science.gov (United States)

    Phillips, T. G.

    1998-05-01

    Simple hydride molecules are of great importance in astrophysics and astrochemistry. Physically they dominate the cooling of dense, warm phases of the ISM, such as the cores and disks of YSOs. Chemically they are often stable end points of chemical reactions, or may represent important intermediate stages of the reaction chains, which can be used to test the validity of the process. Through the efforts of astronomers, physicists, chemists, and laboratory spectroscopists we have an approximate knowledge of the abundance of some of the important species, but a great deal of new effort will be required to achieve the comprehensive and accurate data set needed to determine the energy balance and firmly establish the chemical pathways. Due to the low moment of inertia, the hydrides rotate rapidly and so have their fundamental spectral lines in the submillimeter. Depending on the cloud geometry and temperature profile they may be observed in emission or absorption. Species such as HCl, HF, OH, CH, CH(+) , NH_2, NH_3, H_2O, H_2S, H_3O(+) and even H_3(+) have been detected, but this is just a fraction of the available set. Also, most deduced abundances are not nearly sufficiently well known to draw definitive conclusions about the chemical processes. For example, the most important coolant for many regions, H_2O, has a possible range of deduced abundance of a factor of 1000. The very low submillimeter opacity at the South Pole site will be a significant factor in providing a new capabilty for interstellar hydride spectroscopy. The new species and lines made available in this way will be discussed.

  9. Ultrasonic Inspection Technique and Qualification for Reactor Pressure Vessel Weld of Nuclear Power Plant%核电站反应堆压力容器焊缝的超声检测及验证

    Institute of Scientific and Technical Information of China (English)

    许远欢; 聂勇

    2013-01-01

      The ultrasonic inspection technique for reactor pressure vessel weld of nuclear power plant was described.The ultrasonic transmission characterization and the influence factors of defects detection and sizing are analyzed in detail .A serial of testing is done to verify the UT technique .The ultrasonic tech-nique can effectively detect and size the defects and has been qualified by the separate qualification cen -ter UK-IVC.The technique meets the requirement of ultrasonic inspection of in -service inspection rules for the mechanical components of PWR nuclear islands (RES-M 1997).%  通过对反应堆压力容器焊缝超声波传播特性以及超声波缺陷探测和定量影响因素的分析,并通过大量的试验测试研究,确定了反应堆压力容器焊缝超声检测技术。此反应堆压力容器超声检测技术,能有效地进行缺陷探测和缺陷定量,并通过了英国验证中心的第三方独立验证,满足核电站役前和在役检查规范(RSE-M 1997)的超声检测技术要求。

  10. Smelting Process of SA-508-3-1 Steel for Nuclear Plant Reactor Pressure Vessel%核电压力容器用 SA-508-3-1钢的冶炼

    Institute of Scientific and Technical Information of China (English)

    薛永栋; 晋帅勇; 汪勇; 郭彪

    2012-01-01

      The SA-508-3-1 ingot steel smelting process of EBT smelting-LF refining-VD-VC has been put forward by analyzing the difficulties of smelting process based on the characteristics of steel structure and property requirement for nuclear plant reactor pressure vessel .The nuclear power SA-508-3-1 ingot steel has been produced successfully ac-cording to this steel smelting process .The property of the forgings has reached the requirement of NRPV forgings after forging and heat treatment processes which helped CITIC HEAVY INDUSTRIES to be certified by National Nuclear Safety Administration.%  针对核电压力容器用SA-508-3-1钢的组织与性能特点,分析出冶炼的难点,通过对冶炼工艺的深入研究,提出了采用EBT初炼—LF精炼—真空脱气—真空浇注的冶炼工艺方案,按该工艺方案成功冶炼浇注了核电SA-508-3-1钢锭。经过锻造及热处理等工序后,锻件性能达到核电压力容器锻件要求,并获得国家核安全局认证。

  11. Brief Analysis of Factors Influencing Installing of Reactor Pressure Vessel Stud on Site%浅析影响反应堆压力容器主螺栓现场安装的因素

    Institute of Scientific and Technical Information of China (English)

    胡大芬; 杨景超; 刘东杰; 陈莲

    2016-01-01

    Based on occurred CPR1000 reactor pressure vessel stud unscrewing in stud hole on site by u-sing multi stud tensioning machine(MSTM)in a domestic nuclear power plant,and combined the charac-teristics of its′stud screw thread structure and performance of MSTM,the influencing factors causing this phenomena were analyzed.The influencing mechanism of each factor was discussed.Then corresponding preventive measures were proposed.%针对国内某核电厂 CPR1000反应堆压力容器主螺栓采用整体螺栓拉伸机现场安装时无法顺利旋入主螺孔的现象,结合其主螺栓螺纹副结构和整体螺栓拉伸机性能特点,分析了造成该现象的影响因素,讨论了各因素的影响机理,并提出了相应的预防措施。

  12. Ultrasonic and eddy current testing of austenitic platings of reactor pressure vessels - qualification according the ENIQ method; Ultraschall- und Wirbelstrompruefung austenitischer Plattierungen von Reaktorbehaeltern - Qualifizierung nach der Methode von ENIQ

    Energy Technology Data Exchange (ETDEWEB)

    Just, T.; Csapo, G. [TUeV NORD SysTec GmbH and Co. KG, Hamburg (Germany); Brenner, W. [TUeV Sueddeutschland-ET, Mannheim (Germany); Waidele, H. [MPA, Univ. Stuttgart (Germany)

    2004-07-01

    In the context of the research project SR2318, which received funds from the Federal Radiation Protection Office (Bundesamt fuer Strahlenschutz) and the BMU, the accuracy of eddy current and ultrasonic tests of austenitic platings on reactor pressure vessels was investigated. The results were evaluated with a view to qualification of combined ultrasonic and eddy current tests of platings and base materials on the one hand; on the other hand, a standard test procedure according to the ENIQ method (European Network of Inspection Qualification) is proposed which can serve as a basis for qualification of test procedures. Summarizing suggestions are made for updating the KTA 3201.4 regulation for recurrent inspections of platings. The results of research project SR 2351 are considered. (orig.) [German] Im Rahmen des vom Bundesamt fuer Strahlenschutz bzw. des BMU gefoerderten Untersuchungsvorhabens SR2318 wurde die Aussagefaehigkeit von Ultraschall- und Wirbelstrompruefungen an austenitischen Plattierungen von Reaktordruckbehaeltern untersucht. Die im Untersuchungsvorhaben erzielten Ergebnisse werden zum einen hinsichtlich der Qualifizierung der Pruefung von Plattierungen und des daran angrenzenden Grundwerkstoffes mittels mechanisierter Ultraschallpruefung in Kombination mit der Wirbelstrompruefung bewertet, und es wird zum anderen beispielhaft ein Muster fuer eine Qualifizierung nach der Methodik von ENIQ (European Network of Inspection Qualification) vorgeschlagen, nach dem kuenftig bei Qualifikationen von Pruefverfahren vorgegangen werden kann. Als Quintessenz werden daraus Vorschlaege zur Ergaenzung der Regel KTA 3201.4 hinsichtlich der wiederkehrenden Pruefungen (WKP) von Plattierungsbereichen formuliert. Dabei sind die Ergebnisse des Untersuchungsvorhabens SR 2351 einbezogen worden. (orig.)

  13. The Determination of Reactor Vessel Downcomer Dimensions in PWR Scaling Tests%压水堆比例试验中反应堆压力容器下降段宽度的确定

    Institute of Scientific and Technical Information of China (English)

    王含; 李玉全; 叶子申; 陈炼

    2012-01-01

    This paper presents the reactor vessel downcomer scaling analysis and scaling criteria for the scaled test facility, based on the Hierarchical Two -Tiered Scaling (H2TS) Methodology. Four different methods for the determination of the downcomer dimensions were investigated and discussed for simulating the important thermal - hydraulic phenomena in the downcomer. In the engineering design of the integral test facilities, multiple considerations of these methods should be carefully evaluated.%本文采用H2TS比例分析方法对比例试验中压力容器下降环腔的宽度进行了比例分析,得到了相似设计准则,分析表明为了能够准确的模拟下降段重要的物理现象需要考虑四个与宽度设计相关的相似准则,在设计不同比例的整体性试验台架时,需要综合考虑各现象的相似比例设计要求.

  14. Development of a metal hydride refrigeration system as an exhaust gas-driven automobile air conditioner

    Energy Technology Data Exchange (ETDEWEB)

    Qin, Feng; Chen, Jiangping; Chen, Zhijiu [Institute of Refrigeration and Cryogenics Engineering, Shanghai Jiaotong University, Shanghai 200030 (China); Lu, Manqi; Yang, Ke [Engineering Center, Institute of Metal Research, Chinese Academy of Sciences, Shenyang, Liaoning Province 110016 (China); Zhou, Yimin [Research Center, Zhejiang Yinlun Machiner